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

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.

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

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

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

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

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

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.)

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.

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)

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).

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.

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

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

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.)

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)

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)

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)

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

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.)

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.)

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.

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.

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)

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

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

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

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.

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.

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)

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

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

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

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)

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

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

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

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.)

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

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

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

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

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

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.

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

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

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

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

Development for analysis system of rods enrichment of nuclear fuels

International Nuclear Information System (INIS)

Rojas C, E.L.

1998-01-01

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

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

Visualization test facility of nuclear fuel rod emergency cooling system

International Nuclear Information System (INIS)

Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

2013-01-01

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

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.

Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

International Nuclear Information System (INIS)

Lima, R. de C.F. de; Carajilescov, P.

1987-01-01

The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

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

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

Method and apparatus for sizing nuclear fuel rod cladding tubes

International Nuclear Information System (INIS)

Koehler, L.

1976-01-01

Nuclear fuel rod cladding tubes are sized internally to diameters precisely fitting nuclear fuel pellets with which the tubes are charged by externally applying hydraulic pressure to short lengths of each tube. The pressure is applied while the tube is stationary. The tube is then moved to bring a new length within the hydraulic pressure zone. The volume of the hydraulic liquid used and the pressure applied to this liquid is such that the liquid is compressed slightly so that the length being sized yields, the expansion of the liquid then completing the sizing. The lengths being sized step-by-step are internally supported by either the fuel pellets or a mandrel having the same diameter as the pellets

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.)

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)

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

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.

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

Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

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)

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)

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

Fuel rod for liquid metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Vinz, P.

1976-01-01

In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

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

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

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)

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

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

ORNL capability to conduct post irradiation examination of full-length commercial nuclear fuel rods

International Nuclear Information System (INIS)

Spellman, Donald J.

2007-01-01

Hot cells at Oak Ridge National Laboratory (ORNL) are nearing completion of a multi-year upgrade program to implement 21. century capabilities to meet the examination demands for higher burnup fuels and the future demands that will come from fuel recycling programs. Fuel reliability and zero tolerance for fuel failure is more than an industry goal. Fuel reliability is becoming a requirement that supports the renaissance of nuclear power generation. Thus, fuel development and management of new forms of waste that will come from programs such as the Global Nuclear Energy Partnership (GNEP) will require extensive use of the flexible, high-quality, technically advanced hot cells at ORNL. ORNL has the capability to perform post irradiation examination (PIE) of irradiated commercial nuclear fuel rods and the management structure to ensure a timely, cost-effective result. ORNL can: 1) Handle the transportation issues, 2) Perform macroscopic fuel rod examinations, 3) Perform microscopic fuel and clad examinations, and 4) Manage legacy material and waste disposal issues from PIE activities. All four of these items will be managed in a way that allows the customer day-to-day access to the results and data. Hot cell examination equipment that is necessary to determine the characteristics and performance of irradiated materials must operate in a hostile environment and is subject to long-term degradation that may result in reliability and quality assurance (QA) issues. ORNL has modernized its hot cell nuclear fuel examination equipment, installing state-of-the-art automated examination equipment and data gathering capabilities. ORNL is planning a major commitment to nuclear fuel examination and development, and future improvements will continue to be made over the next few years. (author)

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.)

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

Science.gov (United States)

Mariani, Robert Dominick

2014-09-09

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

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

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)

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

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

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.)

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

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

Procyon 1. First prototype worldwide for storage spent nuclear fuel rods

International Nuclear Information System (INIS)

Meyering, Manfred

2010-01-01

HFH Herbst has designed and built a unique machine for storage of spent highly radioactive nuclear fuel rods within two years for the Swedish SKB. The vehicle (total weight 98 t) can be operated underground without a driver. Herbst was able to bring to this project almost 30 years of experience in the complementation of vehicle projects for the nuclear industry. The Procyon 1 already proved its efficiency impressively in several hundred storage processes and operates with absolute reliability. (orig.)

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

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

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.

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)

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.)

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

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

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.

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)

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)

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.)

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)

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

International Nuclear Information System (INIS)

Johnson, G.L.

1988-09-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

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)

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.

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

Method and system for the production of fuel rods for nuclear reactors

International Nuclear Information System (INIS)

Meinhardt, V.; Schultz, R.; Gall, A.

1975-01-01

The invention deals with a method to produce nuclear reactor fuel rods which contain the fuel in a cladding tube and in which an alkali metal is put in the space between the fuel and cladding for better heat transfer from the fuel to the cladding. The bubble-free filling-in of the alkali metal causes difficulties; the invention suggests to carry out the filling-in of fuel and alkali metal in a horizontal position of the cladding to rinse through the free space firstly with alkali metal, then to fill with fuel and then with the alkali metal taking account of necessary space for the fission gas volume. Equipment to carry out this process in which all operations are carried out in a vacuum box is described. (UWI) [de

Welding nuclear reactor fuel rod end plugs

International Nuclear Information System (INIS)

Yeo, D.

1984-01-01

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

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1984-01-01

The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

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

The effects of off-center pellets on the temperature distribution and the heat flux distribution of fuel rods in nuclear reactors

International Nuclear Information System (INIS)

Peng Muzhang; Xing Jianhua

1986-01-01

This paper analyzes the effects of off-center pellets on the steady state temperature distribution and heat flux distribution of fuel rods in the nuclear reactors, and derives the dimensionless temperature distribution relationships and the dimensionless heat flux distribution relationship from the fuel rods with off-center pellets. The calculated results show that the effects of off-center will result in not only deviations of the highest temperature placement in the fuel pellets, but also the circumferentially nonuniform distributions of the temperatures and heat fluxes of the fuel rod surfaces

Automatic system of welding for nuclear fuel rods

International Nuclear Information System (INIS)

Romero G, M; Romero C, J.

1998-01-01

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

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

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

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

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)

Method of manufacturing nuclear fuel rods

International Nuclear Information System (INIS)

Sato, Masao; Oyama, Masatoshi; Yamamoto, Takanobu.

1976-01-01

Object: To discriminate the properties of light white deposits on a clad tube during the process of manufacturing nuclear fuel rods and then remove this to reproduce a good clad tube, thereby enhancing a yield of the clad tube. Structure: When a light white deposits is found to be appeared on outer or inner surface of coating during the process of appearance inspection, this is then permitted to subject to treatment of hot water immersion and discrimination. Requirements for removal of adhered matter in the process of treatment of hot water immersion are that deioned water of specific resistance 5 x 10 5 ohms or more is used with water temperature maintained at 60 to 100 0 C for immersion treatment for 10 to 30 minutes. In this case, however, if the water temperature is more than 80 0 C, the immersion time can be set less than 10 minutes. With the addition of such process described above, about 2.5% of total receiving number can be reproduced. (Yoshihara, H.)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

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

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

1977-01-01

A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-01

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

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

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

Energy Technology Data Exchange (ETDEWEB)

Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2002-10-01

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). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

International Nuclear Information System (INIS)

Benedetti, R.L.; Lords, L.V.; Kiser, D.M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage

Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods

International Nuclear Information System (INIS)

Jiang Yijie; Cui Yi; Huo Yongzhong; Ding Shurong

2011-01-01

Highlights: → We establish three-dimensional finite element models for nuclear fuel rods. → The thermal-mechanical behaviors at the initial stage of burnup are obtained. → Several parameters on the in-pile performances are investigated. → The parameters have remarkable effects on the in-pile behaviors. → This study lays a foundation for optimal design and irradiation safety. - Abstract: In order to implement numerical simulation of the thermal-mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal-mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal-mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm 3 , the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm 2 K to 0.01 W/mm 2 K, while increases up to 54.7% when h decreases from 0.01 W/mm 2 K to 0.005 W/mm 2 K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal-mechanical behaviors in the fuel rod; when the

Nuclear fuel pellet loading machine

International Nuclear Information System (INIS)

Dazen, J.R.; Denero, J.V.

1976-01-01

A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

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

Gray rod for a nuclear reactor

International Nuclear Information System (INIS)

Francis, T.A.; Cerni, Samuel.

1986-01-01

The invention relates to an improved gray rod for insertion in a nuclear fuel assembly having an array of fuel rods. The gray rod includes a thin-walled cladding tube a first longitudinal section of which is positioned within, and a second longitudinal section of which is positioned essentially without, the array of fuel rods when the gray rod is inserted in the fuel assembly. The first longitudinal section defines a pellet-receiving space having detained therein a stack of annular pellets with an outer diameter sufficient to lend radial support to the wall of the first longitudinal tube section. The second longitudinal section defines a hollow space devoid of pellets and having means to resist radial collapse under external pressure. This means may be a partially compressed spiral spring which serves the dual purpose of retaining the stack of pellets in the pellet-receiving space and of lending radial support to the wall of the second longitudinal tube section or it may be holes through the wall to allow pressure equalisation. The cladding tube is composed of stainless-steel material having a low neutron-capture cross-section, and the annular pellets preferably being composed of Zircaloy or Zirconia material. (author)

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

Nuclear fuels

International Nuclear Information System (INIS)

Gangwani, Saloni; Chakrabortty, Sumita

2011-01-01

Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

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

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

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

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

PWR Fuel licensing in France - from design to reprocessing: licensing of nuclear PWR fuel rod design to satisfy with criteria for normal and abnormal fuel operation in France

International Nuclear Information System (INIS)

Beraha, R.

1999-01-01

In this lecture are presented: French regulatory context; Current fuel management methods; Request from the french operator EdF; Most recent actions of the french Nuclear safety authority; Fuel assemblies deformations (impact of high burn-up; investigations during reactor's exploitation; control rods drop off times)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Fuel assembly for nuclear reactor

International Nuclear Information System (INIS)

Yamanaka, Akihiro; Haikawa, Katsumasa; Haraguchi, Yuko; Nakamura, Mitsuya; Aoyama, Motoo; Koyama, Jun-ichi.

1996-01-01

In a BWR type fuel assembly comprising first fuel rods filled with nuclear fission products and second fuel rods filled with burnable poisons and nuclear fission products, the concentration of the burnable poisons mixed to a portion of the second fuel rods is controlled so that it is reduced at the upper portion and increased at the lower portion in the axial direction. In addition, a product of the difference of an average concentration of burnable poisons between the upper portion and the lower portion and the number of fuel rods is determined to higher than a first set value determined corresponding to the limit value of a maximum linear power density. The sum of the difference of the average concentration of the burnable poisons between the upper portion and the lower portion of the second fuel rod and the number of the second fuel rods is determined to lower than a second set value determined corresponding to a required value of a surplus reactivity. If the number of the fuel rods mixed with the burnable poisons is increased, the infinite multiplication factor at an initial stage of the burning is lowered and, if the concentration of the mixed burnable poisons is increased, the time of exhaustion of the burnable poisons is delayed. As a result, the maximum value of the infinite multiplication factor is suppressed thereby enabling to control surplus reactivity. (N.H.)

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

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

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.

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.)

The study on the methods for improving the gredibility of NDT equipment for the gap of pellets of nuclear fuel rods

International Nuclear Information System (INIS)

Zhang Lei; Liu Ming; Wang Changhong; Ma Jinbo

2014-01-01

In order to improve the credibility of the new generation of automatic online non-destructive testing equipment for the gap of the pellets of nuclear fuel rods the researchers have done a lot of work in the development of the device. Such measures as multi-thread synchronization, precise timing, upper and lower computer communication control, antijamming processing are adopted such that the detecting device can accurately detecte the size of the gap between pellets, the position and length of the spring cavity at the front end of the nuclear fuel rods at a detection rate of 8 m/min. The detection credibility for the 0.5 mm gap is over 95%, reaching the international advanced level. At present, the device is put into use in the nuclear fuel element production line. (authors)

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.)

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.

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

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

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

2015-01-01

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

Encapsulating spent nuclear fuel

International Nuclear Information System (INIS)

Fleischer, L.R.; Gunasekaran, M.

1979-01-01

A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

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.)

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

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

Energy Technology Data Exchange (ETDEWEB)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

Fuel element for nuclear reactors

International Nuclear Information System (INIS)

Cadwell, D.J.

1982-01-01

The invention concerns a fuel element for nuclear reactors with fuel rods and control rod guide tubes, where the control rod guide tubes are provided with flat projections projecting inwards, in the form of local deformations of the guide tube wall, in order to reduce the radial play between the control rod concerned and the guide tube, and to improve control rod movement. This should ensure that wear on the guide tubes is largely prevented which would be caused by lateral vibration of the control rods in the guide tubes, induced by the flow of coolant. (orig.) [de

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

International Nuclear Information System (INIS)

Johnson, G.L.

1991-11-01

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

Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

International Nuclear Information System (INIS)

Larson, J.R.; Evans, D.R.; McCardell, R.K.

1978-01-01

This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.

1982-01-01

A fuel assembly in a nuclear reactor comprises a locking mechanism that is capable of locking the fuel assembly to the core plate of a nuclear reactor to prevent inadvertent movement of the fuel assembly. The locking mechanism comprises a ratchet mechanism 108 that allows the fuel assembly to be easily locked to the core plate but prevents unlocking except when the ratchet is disengaged. The ratchet mechanism is coupled to the locking mechanism by a rotatable guide tube for a control rod or water displacer rod. (author)

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

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

Nuclear fuel element

International Nuclear Information System (INIS)

1974-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Nuclear fuel element

International Nuclear Information System (INIS)

Thompson, J.R.; Rowland, T.C.

1976-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

Ventilating system for reprocessing of nuclear fuel rods

International Nuclear Information System (INIS)

Szulinski, M.J.

1981-01-01

In a nuclear facility such as a reprocessing plant for nuclear fuel rods, the central air cleaner discharging ventilating gas to the atmosphere must meet preselected standards not only as to the momentary concentration of radioactive components, but also as to total quantity per year. In order to comply more satisfactorily with such standards, reprocessing steps are conducted by remote control in a plurality of separate compartments. The air flow for each compartment is regulated so that the air inventory for each compartment has a slow turnover rate of more than a day but less than a year, which slow rate is conveniently designated as quasihermetic sealing. The air inventory in each such compartment is recirculated through a specialized processing unit adapted to cool and/or filter and/or otherwise process the gas. Stale air is withdrawn from such recirculating inventory and fresh air is injected (eg., By the less than perfect sealing of a compartment) into such recirculating inventory so that the air turnover rate is more than a day but less than a year. The amount of air directed through the manifold and duct system from the reprocessing units to the central air cleaner is less than in reprocessing plants of conventional design

Nuclear fuel assembly

International Nuclear Information System (INIS)

Domoto, Noboru; Masuda, Hiroyuki

1989-01-01

In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

Energy Technology Data Exchange (ETDEWEB)

Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

2015-11-15

Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

International Nuclear Information System (INIS)

Tobin, J.G.

2009-01-01

The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

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

Nuclear fuel assembly

International Nuclear Information System (INIS)

Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

1994-01-01

A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

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

Synergistic smart fuel for in-pile nuclear reactor measurements

Energy Technology Data Exchange (ETDEWEB)

Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

2013-07-01

The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

Dispersion fuel for nuclear research facilities

International Nuclear Information System (INIS)

Kushtym, A.V.; Belash, M.M.; Zigunov, V.V.; Slabospitska, O.O.; Zuyok, V.A.

2017-01-01

Designs and process flow sheets for production of nuclear fuel rod elements and assemblies TVS-XD with dispersion composition UO_2+Al are presented. The results of fuel rod thermal calculation applied to Kharkiv subcritical assembly and Kyiv research reactor VVR-M, comparative characteristics of these fuel elements, the results of metallographic analyses and corrosion tests of fuel pellets are given in this paper

Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

Energy Technology Data Exchange (ETDEWEB)

Rojas C, E.L

1998-11-01

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

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

Drying of encapsulated parts (nuclear fuel rods) in applying vacuum, by introducing dehydratings, vacuum, and filling with an inert gas

International Nuclear Information System (INIS)

Johnson, C.R.

1976-01-01

This invention concerns a decontamination technique, in particular a process and equipment for extracting the water contained in fuel rods and other similar components of a nuclear reactor. The extraction of the contaminants contained in the fuel rods is carried out by a standard method by drilling a small hole in the surface of the cladding and applying a vacuum to bleed the rod of its impurities (moisture and gas). The invention consists for example in applying a vacuum at the hole drilled in the cladding to extract the contaminants and introducing spirit into the rod through the same orifice. The spirit absorbs the remaining liquid and other impurities. The spirit charged with the impurities is then pumped out by the same aperture by means of a regulated atmosphere inside a closed receptacle. This receptacle is then filled with an inert gas cooled to ambient temperature. The rods are then pressurised and the small orifice is sealed [fr

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.

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

External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

International Nuclear Information System (INIS)

Welty, R.K.

1980-01-01

The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis

Boosting nuclear fuels

International Nuclear Information System (INIS)

Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

2002-01-01

This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

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

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

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)

Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

Directory of Open Access Journals (Sweden)

Armando C. Marino

2011-01-01

Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

Reconstitutable fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Ferlan, S.J.; Kmonk, S.; Schallenberger, J.M.

1982-01-01

A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

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

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

Assembly mechanism for nuclear fuel bundles

International Nuclear Information System (INIS)

Long, J.W.; Flora, B.S.; Ford, K.L.

1980-01-01

The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

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

Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard

International Nuclear Information System (INIS)

Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M.

1990-01-01

This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab

37-Active rods fuel element for Atucha 1 nuclear power plant. Effects of this change in design over the neutronic behavior, decay power and radioactive inventory

International Nuclear Information System (INIS)

Villar, Javier E.

1999-01-01

The influence of the use of 37-rods fuel element on the behavior of the Atucha 1 nuclear power plant homogeneous core with slightly enriched fuel to 0.85 w % were studied through representative parameters such as average discharge burnup, channel powers, reactivity coefficients, kinetic parameters, radioactive inventory and decay power. In general, the values of mentioned parameters are similar to those corresponding to a core with the 36-rods fuel element actually in use, although it must be emphasized a decrease both in linear power and, in minor degree, in the efficiency of shut-off and control rods and a slight increase in the discharge burnup. The fuel management strategy developed for a core with 36-rods elements can be maintained. (author)

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.

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)

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.)

An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

International Nuclear Information System (INIS)

Costa, Antonio Carlos Lopes da; Machado, Luiz; Koury, Ricardo Nicolau Nassar; Passos, Julio Cesar

2009-01-01

This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

Energy Technology Data Exchange (ETDEWEB)

Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

2009-07-01

This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

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

Fuel element shipping shim for nuclear reactor

International Nuclear Information System (INIS)

Gehri, A.

1975-01-01

A shim is described for use in the transportation of nuclear reactor fuel assemblies. It comprises a member preferably made of low density polyethylene designed to have three-point contact with the fuel rods of a fuel assembly and being of sufficient flexibility to effectively function as a shock absorber. The shim is designed to self-lock in place when associated with the fuel rods. (Official Gazette)

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.

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

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

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)

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

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M.

2013-01-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2013-07-01

In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

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

Cracking and relocation of UO2 fuel during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Dagbjartsson, S.J.

1981-01-01

Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

Energy Technology Data Exchange (ETDEWEB)

James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

2013-05-01

A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

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)

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

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Fuel bundle for nuclear reactor

International Nuclear Information System (INIS)

Long, J.W.; Flora, B.S.; Ford, K.L.

1977-01-01

The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

Science.gov (United States)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

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.

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.

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

Nuclear fuel element end fitting

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

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.)

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)

A partial grid for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Demario, E.E.

1985-01-01

The invention relates to a nuclear-reactor fuel assembly including fuel-rod supporting transverse grids. The fuel assembly includes at least one additional transverse grid which is disposed between two fuel-rod supporting grids and consists of at least one partial grid structure extending across only a portion of the fuel assembly and having fuel rods and control-rod guide thimbles of only said portion extending therethrough. The partial grid structure includes means for providing lateral support of the fuel rods and/or means for laterally deflecting coolant flow, and it is formed of inter-leaved inner straps and border straps, the interleaved inner straps preferably being of substantially smaller height than the border straps to reduce the amount of material capable of parasitically absorbing neutrons. The additional transverse grid may comprise several partial grid structures associated with different groups of fuel rods of the fuel assembly

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

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.

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.

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.

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

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

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

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.

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

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.)

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.

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)

CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

2010-01-01

The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

Control rod for the operation of nuclear reactor

International Nuclear Information System (INIS)

Ishida, Hiromi

1987-01-01

Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

A numerical solution model of the rewetting of a nuclear fuel rod

International Nuclear Information System (INIS)

Braz Filho, F.A.

1984-01-01

The study of thermal behaviour of a nuclear reactor fuel rod during the reflooding phase of the loss-of-coolant accident (LOCA) is presented. A mathematical model and a numerical scheme were proposed in order to solve the bidimensional heat conduction equation in cylindrical coordinates. The phenomenon of reflooding is not completely understood. One of the main difficulties is to estimate the heat transfer coefficient (h). For this reason two different models were elaborated: in the first three regions are considered and in each region h is considered constant; in the second the h profile is adjusted according to the boiling curve. The three region model yields satisfactory results at high and low mass flows while the 'boiling curve' model yields reasonable at low flows. (Author) [pt

Method of manufacturing nuclear fuel pellet

International Nuclear Information System (INIS)

Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

1989-01-01

In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

Prevention of nuclear fuel cladding materials corrosion

International Nuclear Information System (INIS)

Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

1983-01-01

The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

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)

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.

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)

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

Nuclear fuel bundle disassembly and assembly tool

International Nuclear Information System (INIS)

Yates, J.; Long, J.W.

1975-01-01

A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

Utilization of carbon/carbon composites in nuclear simulation fuel rods

International Nuclear Information System (INIS)

Polidoro, H.A.; Otani, S.; Rezende, M.C.; Ferreira, S.R.; Otani, C.

1988-01-01

Thermo-hydraulic problems, in nuclear plants are normally analysed by using electrically heated rods. Carbon/carbon composites were used to make heating elements for testing by indirect heating up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceed if the choice of the complementary materials for insulator and cladding were improved. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. (author) [pt

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

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

Support grid for fuel elements in a nuclear reactor

International Nuclear Information System (INIS)

Finch, L.M.

1977-01-01

A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates

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)

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)

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.

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)

A Study on the Sliding/Impact Wear of a Nuclear Fuel Rod in Room Temperature Air: (I) Development of a Test Rig and Characteristic Analysis

International Nuclear Information System (INIS)

Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2007-01-01

A new type of a fretting wear tester has been designed and developed in order to simulate the actual vibration behavior of a nuclear fuel rod for springs/dimples in room temperature. When considering the actual contact condition between fuel rod and spring/dimple, if fretting wear progress due to the Flow-Induced Vibration (FIV) under a specific normal load exerted on the fuel rod by the elastic deformation of the spring, the contacting force between the fuel rod and dimple that were located in the opposite side should be decreased. Consequently, the evaluation of developed spacer grids against fretting wear damage should be performed with the results of a cell unit experiments because the contacting force is one of the most important variables that influence to the fretting wear mechanism. Therefore, it is necessary to develop a new type of fretting test rig in order to simulate the actual contact condition. In this paper, the development procedure of a new fretting wear tester and its performance were discussed in detail

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)

Dismantling method for nuclear fuel assembly

International Nuclear Information System (INIS)

Yamazaki, Shuji; Kato, Akihiro; Yoshida, Masafumi.

1993-01-01

An upper nozzle is detached from a control rod guide tube and an instrumentation tube. Subsequently, slots (slits) having a predetermined width are formed longitudinally at enlarged diameter portions of the control rod guide tube and the instrumentation tube. Then, the control rod guide tube and the instrumentation tube are separated from a lower nozzle, and pulled out from the lattice space of each of the support lattices. Thereafter, a predetermined key is inserted to a key insertion window formed at each of the support lattices, to distort a spring and take the fuel rod out of the lattice space of each of the support lattices. With such procedures, when the control rod guide tube and the instrumentation tube are pulled out of the lattice space of the support lattice, the enlarged diameter portion is narrowed to reduce the diameter, thereby enabling to take them out easily. Accordingly, since the space for inserting the key can be ensured, the nuclear fuel assemblies can easily be dismantled. In addition, fuel rods can be taken out smoothly and in an intact state. (I.N.)

Quality management of nuclear fuel

International Nuclear Information System (INIS)

2006-01-01

The Guide presents the quality management requirements to be complied with in the procurement, design, manufacture, transport, receipt, storage, handling and operation of nuclear fuel. The Guide also applies to control rods and shield elements to be placed in the reactor. The Guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organisations, whose activities affect fuel quality and the safety of fuel transport, storage and operation. General requirements for nuclear fuel are presented in Section 114 of the Finnish Nuclear Energy Decree and in Section 15 of the Government Decision (395/1991). Regulatory control of the safety of fuel is described in Guides YVL6.1, YVL6.2 and YVL6.3. An overview of the regulatory control of nuclear power plants carried out by STUK (Radiation and Nuclear Safety Authority, Finland) is clarified in Guide YVL1.1

Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

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)

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

Study of correcting the effect of daughter age on determining 235U enrichment of fuel rods

International Nuclear Information System (INIS)

Deng Jingshan; Zhou Chengfang; Luo Minxuan; Liu Yun

1997-01-01

Gamma-ray passive technique is a very effective method to assay and determine 235 U enrichment of nuclear power plant fuel rods. There is a weakness in this passive method, i.e. only after the uranium isotope daughters of UO 2 pellets have reached to equilibrium with uranium parent, then the 235 U enrichment can be determined. This weakness greatly restricts the application of the method. A new two-peak and two-window technique is developed that can overcome the interference of uranium daughter decay in determining 235 U enrichment of nuclear fuel rods, and the results are very satisfactory. The new technique will play an important role in the gamma-ray passive technique for determining 235 U enrichment of fuel rods. This new technique also makes the gamma-ray passive method perfectly. (11 figs., 6 tabs.)

Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

International Nuclear Information System (INIS)

Cho, Byungoh.

1990-01-01

Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

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

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

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

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

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

Control rod cluster with removable rods for nuclear fuel assembly

International Nuclear Information System (INIS)

Denizou, J.P.

1989-01-01

For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod [fr

Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

International Nuclear Information System (INIS)

Zoran Stosic; Vladimir Stevanovic

2005-01-01

Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

Process and device for fastening and removing fuel-absorber rods in fuel elements of nuclear reactors

International Nuclear Information System (INIS)

Edwards, G.T.; Schluderberg, D.C.

1980-01-01

This is concerned with an improvement of the fixing of absorber rods in a nuclear reactor. It is important that the rod should not be damaged during removal from the reactor, and that no particles of material are shed during this process. According to the invention, the rod has a stalk which is pressed into a hole in the star shaped arms and welded in. During removal, the stalk is broken at a preferred position. Details of construction are described. (UWI) [de

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

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)

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

Internal attachment of laser beam welded stainless steel sheathed thermocouples into stainless steel upper end caps in nuclear fuel rods for the LOFT Reactor

International Nuclear Information System (INIS)

Welty, R.K.; Reid, R.D.

1980-01-01

The Exxon Nuclear Company, Inc., acting as a subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, conducted a laser beam welding study to attach internal stainless steel thermocouples into stainless steel upper end caps in nuclear fuel rods. The objective of this study was to determine the feasibility of laser welding a single 0.063 inch diameter stainless steel (304) sheathed thermocouple into a stainless steel (316) upper end cap for nuclear fuel rods. A laser beam was selected because of the extremely high energy input in unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A special weld fixture was designed and fabricated to hold the end cap and the thermocouple with angular and rotational adjustment under the laser beam. A commercial pulsed laser and energy control system was used to make the welds

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

Stress analysis and collapse time prediction of nuclear fuel cladding tube with wear scar

International Nuclear Information System (INIS)

Lee, J. S.; Kim, O. H.; Kim, H. K.; Hu, Y. H.; Kim, J. I.; Kim, K. T.

2004-01-01

In this analysis, the stress and collapse time analysis models for nuclear fuel rod with the fretting wear scar were developed in order to evaluate the effects of the wear depth on the integrity of nuclear fuel rod. The stress analysis result shows that the nuclear fuel rod with approximately 60% deep wear scar of the clad wall thickness, meets the allowable stress criteria and the collapse time analysis indicates that the fuel rod with less than roughly 56% deep wear scar of the clad wall thickness has longer collapse time than the expected fuel life-time. The both stress and collapse time results are evaluated to be very reasonable on considering the comparison with the outputs of existing design code for the simple model. However, the developed analysis models and the results will be confirmed by the tests

Rod consolidation at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Bailey, W.J.

1986-12-01

A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab

Quality assurance monitoring during nuclear fuel production in JSC 'TVEL'

International Nuclear Information System (INIS)

Filimonov, G.; Tchirkov, V.

2000-01-01

The paper describes Quality Assurance (QA) monitoring during fabrication of nuclear fuel in Russian Federation. Joint Stock Company 'TVEL', natural state monopoly of the type of holding that fabricates and supplies nuclear fuel for the NPPs of Russia, CIS and Europe, incorporates the major enterprises of the nuclear fuel cycle including JSC 'Mashinostroitelny zavod', Electrostal (fabrication of fuel pellets, rods and assemblies for different types of reactors), JSC 'Novosibirsky zavod khimconcentratov', Novosibirsk (fabrication of fuel rods and assemblies for WWER-440 and WWER-1000), JSC 'Tchepetsky mechanitchesky zavod', Tchepetsk (fabrication of Zr tubing). Monitoring of QA is an important element of Quality Management System (QMS) developed and implemented at the above-mentioned enterprises of the JSC 'TVEL' and it is performed on three levels including external and internal audits and author's supervision. Paper also describes short- and long-term policies of the JSC 'TVEL' in nuclear fuel quality field. (author)

Conditioning of nuclear reactor fuel

International Nuclear Information System (INIS)

1975-01-01

A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

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

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)

Existing experimental criticality data applicable to nuclear-fuel-transportation systems

International Nuclear Information System (INIS)

Bierman, S.R.

1983-02-01

Analytical techniques are generally relied upon in making criticality evaluations involving nuclear material outside reactors. For these evaluations to be accepted the calculations must be validated by comparison with experimental data for a known set of conditions having physical and neutronic characteristics similar to those conditions being evaluated analytically. The purpose of this report is to identify those existing experimental data that are suitable for use in verifying criticality calculations on nuclear fuel transportation systems. In addition, near term needs for additional data in this area are identified. Of the considerable amount of criticality data currently existing, that are applicable to non-reactor systems, those particularly suitable for use in support of nuclear material transportation systems have been identified and catalogued into the following groups: (1) critical assemblies of fuel rods in water; (2) critical assemblies of fuel rods in water containing soluble neutron absorbers; (3) critical assemblies containing solid neutron absorber; (4) critical assemblies of fuel rods in water with heavy metal reflectors; and (5) critical assemblies of fuel rods in water with irregular features. A listing of the current near term needs for additional data in each of the groups has been developed for future use in planning criticality research in support of nuclear fuel transportation systems. The criticality experiments needed to provide these data are briefly described and identified according to priority and relative cost of performing the experiments

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)

The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

International Nuclear Information System (INIS)

Jung, Jae Hoo; Park, B. S.; Kim, K. H.; Hong, D. H.; Yoon, J. S.; Lee, H. J.; Kim, S. H.; Song, T. G.; Lim, K. M.; Lee, J. K.

2006-12-01

Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO 2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

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

Development of the method for the dimensional measurement of the HANARO nuclear fuel

International Nuclear Information System (INIS)

Kim, Tae Yeon; Lee, K. S.; Park, D. G.; Choo, Y. S.; Ahn, S. B.

1998-06-01

Dimension of the nuclear fuel is altered in nuclear reactor because of the neutron exposure with high pressure water. If the deformation is overlarge, the severe problem in safety of the nuclear fuel and the reactor come about. Therefore the accurate dimensional data of the nuclear fuel in diameter and length is very important for the design of the nuclear fuel and the estimation of the nuclear safety. Measurement of diameter for the dummy HANARO fuel rod which has not filled with real fuel material was carried out in hot cell. And also the length of the HANARO fuel assembly and the rod are measured. Dimensional measuring method for the HANARO fuel was developed. The test result show our method is good enough to distinguish change in volume with statistical uncertainty of 0.6 %. (author). 2 refs., 7 tabs., 20 figs

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)

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

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

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

Automatic system of welding for nuclear fuel rods; Sistema automatico de soldadura para barras de combustible nuclear

Energy Technology Data Exchange (ETDEWEB)

Romero G, M; Romero C, J [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

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

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)

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

The nuclear fuel cycle

International Nuclear Information System (INIS)

Patarin, L.

2002-01-01

This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

Development of TIG Welding System for a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

Development of TIG Welding System for a Nuclear Fuel Test Rig

International Nuclear Information System (INIS)

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

2013-01-01

The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

Experience feedback of operational events of the control rod assembly and its drive mechanism in nuclear power plants

International Nuclear Information System (INIS)

Zhou Hong; Xiao Zhi; Tao Shusheng; Zheng Lixin; Chen Zhaolin

2013-01-01

Seventeen operational events of the control rod assembly and its drive mechanism are collected from 1992 to 2012 important nuclear operational events and feedback in referred nuclear power plants. After investigated and classified, several important issues, such as the impact of control rod swell and fuel assembly distortion, control rod drive mechanism leakage, and the control system reliability of control rod, are emphatically analyzed. Some suggestions of experience feedback are proposed. (authors)

Nuclear fuel pellet charging device

International Nuclear Information System (INIS)

Komuro, Kojiro.

1990-01-01

The present invention concerns a nuclear fuel pellet loading device, in which nuclear fuel pellets are successively charged from an open end of a fuel can while rotating the can. That is, a fuel can sealed at one end with an end plug and opened at the other end is rotated around its pipe axis as the center on a rotationally diriving table. During rotation of the fuel can, nuclear fuel pellets are successively charged by means of a feed rod of a feeding device to the inside of the fuel can. The fuel can is rotated while being supported horizontally and the fuel pellets are charged from the open end thereof. Alternatively, the fuel can is rotated while being supported obliquely and the fuel pellets are charged gravitationally into the fuel can. In this way, the damages to the barrier of the fuel can can be reduce. Further, since the fuel pellets can be charged gravitationally by rotating the fuel can while being supported obliquely, the damages to the barrier can be reduced remarkably. (I.S.)

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.

Contribution to numerical and mechanical modelling of pellet-cladding interaction in nuclear reactor fuel rod

International Nuclear Information System (INIS)

Retel, V.

2002-12-01

Pressurised water reactor fuel rods (PWR) are the place of nuclear fission, resulting in unstable and radioactive elements. Today, the mechanical loading on the cladding is harder and harder and is partly due to the fuel pellet movement. Then, the mechanical behaviour of the cladding needs to be simulated with models allowing to assess realistic stress and strain fields for all the running conditions. Besides, the mechanical treatment of the fuel pellet needs to be improved. The study is part of a global way of improving the treatment of pellet-cladding interaction (PCI) in the 1D finite elements EDF code named CYRANO3. Non-axisymmetrical multidirectional effects have to be accounted for in a context of unidirectional axisymmetrical finite elements. The aim of this work is double. Firstly a model simulating the effect of stress concentration on the cladding, due to the opening of the radial cracks of fuel, had been added in the code. Then, the fragmented state of fuel material has been taken into account in the thermomechanical calculation, through a model which led the strain and stress relaxation in the pellet due to the fragmentation, be simulated. This model has been implemented in the code for two types of fuel behaviour: elastic and viscoplastic. (author)

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)

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

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)

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)

Quality assurance of nuclear fuel

International Nuclear Information System (INIS)

1994-01-01

The guide presents the quality assurance requirements to be completed with in the procurement, design, manufacture, transport, handling and operation of the nuclear fuel. The guide also applies to the procurement of the control rods and the shield elements to be placed in the reactor. The guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organizations whose activities affect fuel quality, the safety of fuel transport, storage and operation. (2 refs.)

Technical verification of advanced nuclear fuel for KSNPs

International Nuclear Information System (INIS)

Lee, C. B.; Bang, J. G.; Kim, D. H. and others

2002-03-01

KNFC has developed the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants through the three-year R and D project (from April 1999 to March 2002) under the Nuclear R and D program by MOST. The purpose of this project is to verify the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants being developed by KNFC during the same period. Verification tests for the advanced fuel assembly and its components such as characteristic test on the spacer grid spring and dimple, static buckling and dynamic impact test on the 5x5 partial spacer grid, the fuel rod vibration test supported by the PLUS7 mid-spacer grid, fretting wear test, turbulent flow structure test in wind tunnel and corrosion test were performed by using the KAERI facilities. Design reports and test results produced by KNFC were technically reviewed. For the domestic production of burnable poison rod, manufacturing technology of burnable poison pellets was developed

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

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

Device for refueling a nuclear reactor having a core comprising a plurality of fuel assemblies

International Nuclear Information System (INIS)

Van Santen, A.; Elofsson, K.

1975-01-01

A nuclear reactor formed of fuel assemblies each including a plurality of parallel fuel rods arranged in a predetermined fuel rod lattice, which rods are freely extractable and insertable at one end of the fuel assembly, is refueled by extracting from one of the fuel assemblies a number of fuel rods substantially less than the total number of fuel rods and replacing these by inserting new fuel rods into the vacated positions. The removal and return of the rods is produced by a tool having a plurality of gripping members capable of engaging shoulders beneath heads formed on the upper ends of the fuel rods. This may be accomplished by providing a tool having a number of gripping members attached to the tool body corresponding to the lattice positions of the fuel rods to be extracted, having gripping members which can be pushed together to grip beneath shoulders on the upper ends of the fuel rods. (Official Gazette)

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)

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)

Gas sealing welding method and device for nuclear fuel rod

International Nuclear Information System (INIS)

Seki, Masayuki; Nishiyama, Motokuni; Kamimura, Katsuichiro; Yagi, Eiji; Nakase, Tsuyoshi; Kobogata, Sadao; Taniguchi, Jun-ichi; Uesugi, Yoshisaku.

1995-01-01

An end plug and a cladding tube are held by clamping, respectively, by opposing movable electrode and static electrode. The movable electrode is forwarded toward the static electrode. The end plug and the cladding tube are abutted and held at a slight gap between their end faces. A region to be welded is surrounded by a pressurizing chamber and the side of the chamber is evacuated and He gas is filled in the cladding tube. Then, one of the electrodes is forwarded, to seal the abutted end faces of the end plug and the cladding tube. Then, pressure and welding current required for welding are applied to the abutted ends, and He gas is sealed in the vessel. The displacement of pressurization caused by slipping when the required pressure is applied to the abutted ends is detected by a sensor, and the operation of the welding control device for starting current supply is terminated by the detection signals. Abutment accuracy between the abutment of the cladding tube and the end plug as a nuclear fuel rod can be ensured, to further improve and stabilize the welding quality. (N.H.)

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

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