Failed fuel detector

International Nuclear Information System (INIS)

Onodera, Koichi.

1981-01-01

Purpose: To improve the reliability of detecting the failure of a fuel rod by imparting a wire disconnection detecting function to a central electrode at the center of a failure mode thereto. Constitution: A wire disconnection detecting terminal is provided at the terminal opposite to the signal output terminal of a central electrode in a failed fuel detector used for detecting the failure of a fuel rod in an atomic power plant using liquid metal as a coolant, and a voltage monitor for monitoring the terminal voltage is connected to the terminal. The disconnection of the central electrode is detected by the failure of the output of the voltage monitor, and an alarm is thus generated. (Aizawa, K.)

Failed fuel detector

International Nuclear Information System (INIS)

Kogure, Sumio; Seya, Toru; Watanabe, Masaaki.

1976-01-01

Purpose: To enhance the reliability of a failed fuel detector which detects radioactivity of nuclear fission products leaked out from fuel elements in cooling water. Constitution: Collected specimen is introduced into a separator and co-existing material considered to be an impediment is separated and removed by ion exchange resins, after which this specimen is introduced into a container housing therein a detector to systematically measure radioactivity. Thereby, it is possible to detect a signal lesser in variation in background, and inspection work also becomes simple. (Kawakami, Y.)

Characteristics of axial splits in failed BWR fuel rods

International Nuclear Information System (INIS)

Lysell, G.; Grigoriev, V.

2000-01-01

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

Method of detecting failed fuels

International Nuclear Information System (INIS)

Ishizaki, Hideaki; Suzumura, Takeshi.

1982-01-01

Purpose: To enable the settlement of the temperature of an adequate filling high temperature pure water by detecting the outlet temperature of a high temperature pure water filling tube to a fuel assembly to control the heating of the pure water and detecting the failed fuel due to the sampling of the pure water. Method: A temperature sensor is provided at a water tube connected to a sipping cap for filling high temperature pure water to detect the temperature of the high temperature pure water at the outlet of the tube, and the temperature is confirmed by a temperature indicator. A heater is controlled on the basis of this confirmation, an adequate high temperature pure water is filled in the fuel assembly, and the pure water is replaced with coolant. Then, it is sampled to settle the adequate temperature of the high temperature coolant used for detecting the failure of the fuel assembly. As a result, the sipping effect does not decrease, and the failed fuel can be precisely detected. (Yoshihara, H.)

Development of failed fuel detection system for PWR (III)

International Nuclear Information System (INIS)

Hwang, Churl Kew; Kang, Hee Dong; Jeong, Seung Ho; Cho, Byung Sub; Yoon, Byeong Joo; Yoon, Jae Seong

1987-12-01

Ultrasonic transducers satisfying the conditions for failed fuel rod detection for failed fuel rod detection have been designed and built. And performance tests for them have been carried out. Ultrasonic signal processing units, a manipulator guiding the ultrasonic probe through the fuel assembly lanes and its control units have been constructed. The performance of the system has been verified experimentally to be successful in failed fuel rod detection. (Author)

A method of failed fuel detection

International Nuclear Information System (INIS)

Uchida, Shunsuke; Utamura, Motoaki; Urata, Megumu.

1976-01-01

Object: To keep the coolant fed to a fuel assembly at a level below the temperature of existing coolant to detect a failed fuel with high accuracy without using a heater. Structure: When a coolant in a coolant pool disposed at the upper part of a reactor container is fed by a coolant feed system into a fuel assembly through a cap to fill therewith and exchange while forming a boundary layer between said coolant and the existing coolant, the temperature distribution of the feed coolant is heated by fuel rods so that the upper part is low whereas the lower part is high. Then, the lower coolant is upwardly moved by the agitating action and fission products leaked through a failed opening at the lower part of the fuel assembly and easily extracted by the sampling system. (Yoshino, Y.)

In-core sipping method for the identification of failed fuel assemblies

International Nuclear Information System (INIS)

Wu Zhongwang; Zhang Yajun

2000-01-01

The failed fuel assembly identification system is an important safety system which ensures safe operations of reactor and immediate treatment of failed fuel rod cladding. The system uses an internationally recognized method to identify failed fuel assemblies in a reactor with fuel element cases. The in-core sipping method is customary used to identify failed fuel assemblies during refueling or after fuel rod cladding failure accidents. The test is usually performed after reactor shutdown by taking samples from each fuel element case while the cases are still in their original core positions. The sample activity is then measured to identify failed fuel assemblies. A failed fuel assembly identification system was designed for the NHR-200 based on the properties of the NHR-200 and national requirements. the design provides an internationally recognized level of safety to ensure the safety of NHR-200

Failed fuel detection device

International Nuclear Information System (INIS)

Doi, Akira.

1994-01-01

The device of the present invention concerns a failed fuel detection device for a nuclear reactor, such as an FBR type reactor, using electroconductive coolants. A sampling port is disposed at the upper portion of the fuel assembly so as to cover the assembly, so that coolants in the fuel assembly are sampled to improve a device for detecting fuel failure. That is, when coolants in the fuel assembly are sampled from the sampling port, the flow of electroconductive coolants in an sampling tube is detected by a flowmeter, to control an electromagnetic pump. The flow of electroconductive coolants is stopped against the waterhead pressure and dynamic pressure of the conductive coolants, and a predetermined amount of the coolants is pumped up to the sampling tank. Gas is supplied to the pumped up coolants so that fissile products are transferred from the coolants to a gas phase. Radiation in the gas in a gas recycling system is measured to detect presence of fuel failure. (I.S.)

Experience with failed or damaged spent fuel and its impacts on handling

International Nuclear Information System (INIS)

Bailey, W.J.

1989-12-01

Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

Method for detecting a failed fuel

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke.

1976-01-01

Purpose: To provide a method for the detection of failed fuel by pouring hot water, in which pouring speed of liquid to be poured and temperature of the liquid are controlled to prevent the leakage of the liquid. Constitution: The method comprises blocking the top of a fuel assembly arranged in coolant to stop a flow of coolant, pouring a liquid higher in temperature than that of coolant into the fuel assembly, sampling the liquid poured, and measuring the concentration of radioactivity of coolant already subjected to sampling to detect a failed fuel. At this time, controlling is made so that the pouring speed of the poured liquid is set to about 25 l/min, and an increased portion of temperature from the temperature of liquid to the temperature of coolant is set to a level less than about 15 0 C. (Furukawa, Y.)

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)

Method of locating a leaking fuel element in a fast breeder power reactor

International Nuclear Information System (INIS)

Honekamp, J.R.; Fryer, R.M.

1976-01-01

Failed fuel detection and location in an FBR is accomplished by mass spectrometric analysis of the cover gas for the 134 Xe/ 133 Xe ratio, and correlation with the theoretical ratio in fuel elements of known power level and burnup. (E.C.B.)

Online failed fuel identification using delayed neutron detector signals in pool type reactors

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

2011-01-01

In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

Failed fuel detection device

International Nuclear Information System (INIS)

Kawai, Masayoshi; Hayashida, Yoshihisa; Niidome, Jiro.

1985-01-01

Purpose: To prevent intrusion of background neutrons to neutron detectors thereby improve the S/N ratio of the detectors in the failed fuel detection device of LMFBR type reactors. Constitution: Neutrons from the reactor core pass through the gaps around the penetration holes in which the primary pipeways pass through the concrete shielding walls and pass through the gaps between the thermal shielding members and the neutron moderating shielding members of the failed fuel detection device and then intrude into the neutron detectors. In view of the above, inner neutron moderating shielding members and movable or resilient neutron shielding members are disposed to the inside of the neutron moderating shielding member. Graphite or carbon hydrides such as paraffin or synthetic resin with a large neutron moderation effect are used as the outer moderating shielding member and materials such as boron or carbon are used for the inner members. As a result, the background neutrons are shielded by the inner neutron moderating shielding members and the resilient neutron shielding members, by which the S/N ratio of the neutron detectors can be increased to 2 - 4 times. (Moriyama, K.)

Detection device for the failed position in fuels

International Nuclear Information System (INIS)

Tokunaga, Kensuke; Nomura, Teiji; Hiruta, Koji

1985-01-01

Purpose: To detect the failed position of a fuel assembly with ease and safety. Constitution: A fuel assembly is tightly closed in a sipper tube equipped with a gas supply tube and a gas exhaust tube at the upper portion and a purified water injection tube and a draining tube at the lower end. Then, water in the sipper tube is drained to the lower portion of the fuel assembly by the pressure of gases while opening the gas supply tube and the draining tube, and closing the exhaust tube and the injection tube. Then, after closing the gas supply tube and the draining tube while opening theexhaust tube and the injection tube, purified water is injected into the sipper tube from the injection tube to an optional height till the fuel assembly is immersed. Then, after leaving for a predetermined of time, water is sampled and the radioactive material density therein is measured. By changing the injection level of the purified water, since the radioactive material density changes at the failed position, the failed position can be detected with ease. (Sekiya, K.)

Fuel failure detection and location in LMFBRs

International Nuclear Information System (INIS)

Jacobi, S.

1982-06-01

The Specialists' Meeting on 'Fuel Failure Detection and Location in LMFBRs' was held at the Kernforschungszentrum Karlsruhe, Federal Republic of Germany, on 11-14 May 1981. The meeting was sponsored by the International Atomic Energy Agency (IAEA) on the recommendation of the International Working Group on Fast Reactors (IWGFR).The purpose of the meeting was to review and discuss methods and experience in the detection and location of failed fuel elements and to recommend future development. The technical sessions were divided into five topical sessions as follows: 1. Reactor Intrumentation, 2. Experience Gained from LMFBRs, 3. In-pile Experiments, 4. Models and Codes, 5. Future Programs. During the meeting papers were presented by the participants on behalf of their countries or organizations. Each presentation was followed by an open discussion in the subject covered by the presentation. After the formal sessions were completed, a final discussion session was held and general conclusions and recommendationswere reached. Session summaries, general conclusions and recommendations, the agenda of the meeting and the list of participants are given. (orig./RW)

Detection method of a failed fuel

International Nuclear Information System (INIS)

Urata, Megumu; Uchida, Shunsuke; Utamura, Motoaki.

1976-01-01

Object: To divide a tank arrangement into a heating tank for the exclusive use of heating and a mixing tank for the exclusive use of mixing to thereby minimize the purifying amount of reactor water pumped from the interior of reactor and to considerably minimize the capacity of a purifier. Structure: In a detection method of a failed fuel comprising stopping a flow of coolant within fuel assemblies arranged in the coolant in a reactor container, sampling said coolant within the fuel assemblies, and detecting a radioactivity level of sampling liquid, the improvement of the method comprising the steps of heating a part of said coolant removed from the interior of said reactor container, mixing said heated coolant into the remainder of said removed coolant, pouring said mixed liquid into said fuel assemblies, and after a lapse of given time, sampling the liquid poured into said fuel assemblies. (Kawakami, Y.)

Detector for failed fuel elements

International Nuclear Information System (INIS)

Ito, Masaru.

1979-01-01

Purpose: To provide automatic monitor for the separation or reactor water and sampling water, in a failed fuel element detector using a sipping chamber. Constitution: A positional detector for the exact mounting of a sipping chamber on a channel box and a level detector for the detection of complete discharge of cooling water in the sipping chamber are provided in the sipping chamber. The positional detector is contacted to the upper end of the channel box and operated when the sipping chamber is correctly mounted to the fuel assemblies. The level detector comprises a float and a limit switch and it is operated when the water in the sipping chamber is discharged by a predetermined amount. Isolation of reactor water and sampling water are automatically monitored by the signal from these two detectors. (Ikeda, J.)

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)

Radiolysis effects on fuel corrosion within a failed nuclear waste container

International Nuclear Information System (INIS)

Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

2003-01-01

The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

Transportation of failed or damaged foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

1999-01-01

Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

The cost of operating with failed fuel at Virginia power

International Nuclear Information System (INIS)

Ford, C.A.

1988-01-01

Virginia Power has completed a study of the costs incurred due to fuel failures in its pressurized water reactors. This study was prompted by histories of high primary coolant activity and subsequent fuel inspections at the North Anna and Surry power stations. The study included an evaluation of the total costs of fuel failures as well as an evaluation of the economics of postirradiation fuel inspections. The major costs of fuel failures included personnel radiation exposure, permanently discharged failed fuel, radwaste generation, increased labor requirements, containment entry delays due to airborne radioactivity, and ramp rate restrictions. Although fuel failures affect a utility in several other areas, the items evaluated in the study were thought to be the most significant of the costs. The study indicated that performing a postirradiation failed fuel examination can be economically justified at tramp-corrected 131 I levels of > 0.015 μCi/g. The savings to the utility can be on the order of several million dollars. Additionally, the cost penalty of performing a fuel inspection at lower iodine levels is generally in the range of $200,000. This economic penalty is expected to be outweighed by the intangible benefits of operating with a defect-free core

Transportation of failed or damaged foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, C.E.; Mustin, T.P.; Massey, C.D.

1998-01-01

Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

2002-01-01

This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

Method and device for detecting failed fuels

International Nuclear Information System (INIS)

Saito, Shozo; Suzumura, Takeshi.

1981-01-01

Purpose: To shorten the time required for inspecting a failed fuel by providing a first outlet for exhausting cleaning liquid to a sampling pipe and a second outlet for exhausting sampled coolant, thereby safely setting a collecting means to the first outlet. Constitution: A sampling pipe is inserted into a fuel assembly loaded within a reactor core, and coolant flow is thus prevented from passing through the fuel assembly interior. Then, with the coolant flow stopped, it is allowed to stand for a predetermined time. Subsequently, cleaning liquid is supplied into the sampling tube and the interior of the sampling pipe is cleaned. Thereafter, the sampling liquid that was in the sampling pipe is exhausted from the first outlet of the sampling pipe. Then, the coolant in the fuel assembly is supplied from the second outlet of the sampling pipe to a collecting means. (Aizawa, K.)

Field experience with failed-fuel detection - PWRs [pressurized water reactors

International Nuclear Information System (INIS)

File, P.

1990-01-01

Industry sensitivity to in-reactor fuel reliability has been heightened recently for several reasons: increased Institute of Nuclear Power Operations attention (fuel reliability indicator), impact of fuel failures on operating and maintenance costs, and concern about personnel contaminations caused by small radioactive particles (often fragments from failed rods). As a result, utilities often perform inspection and repair campaigns at the end of a cycle where fuel failures have been evident in order to avoid experiencing the effects of those failures during the subsequent cycle. At Calvert Cliffs, the first full-core inspection associated with concern over fuel integrity occurred in early 1984 after unit 2 cycle 5. Lessons learned from this campaign are discussed in this paper. While significant improvement has been made relative to the ability to discern failed rods from sound rods, ultrasonic testing should be evaluated by people experienced with mechanisms that can produce questionable indications. The speed and accuracy of UT systems make UT a practical tool for performing inspections with minimal impact on the schedule

Aviation Fuel System Reliability and Fail-Safety Analysis. Promising Alternative Ways for Improving the Fuel System Reliability

Directory of Open Access Journals (Sweden)

I. S. Shumilov

2017-01-01

Full Text Available The paper deals with design requirements for an aviation fuel system (AFS, AFS basic design requirements, reliability, and design precautions to avoid AFS failure. Compares the reliability and fail-safety of AFS and aircraft hydraulic system (AHS, considers the promising alternative ways to raise reliability of fuel systems, as well as elaborates recommendations to improve reliability of the pipeline system components and pipeline systems, in general, based on the selection of design solutions.It is extremely advisable to design the AFS and AHS in accordance with Aviation Regulations АП25 and Accident Prevention Guidelines, ICAO (International Civil Aviation Association, which will reduce risk of emergency situations, and in some cases even avoid heavy disasters.ATS and AHS designs should be based on the uniform principles to ensure the highest reliability and safety. However, currently, this principle is not enough kept, and AFS looses in reliability and fail-safety as compared with AHS. When there are the examined failures (single and their combinations the guidelines to ensure the AFS efficiency should be the same as those of norm-adopted in the Regulations АП25 for AHS. This will significantly increase reliability and fail-safety of the fuel systems and aircraft flights, in general, despite a slight increase in AFS mass.The proposed improvements through the use of components redundancy of the fuel system will greatly raise reliability of the fuel system of a passenger aircraft, which will, without serious consequences for the flight, withstand up to 2 failures, its reliability and fail-safety design will be similar to those of the AHS, however, above improvement measures will lead to a slightly increasing total mass of the fuel system.It is advisable to set a second pump on the engine in parallel with the first one. It will run in case the first one fails for some reasons. The second pump, like the first pump, can be driven from the

Categorization of failed and damaged spent LWR [light-water reactor] fuel currently in storage

International Nuclear Information System (INIS)

Bailey, W.J.

1987-11-01

The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs

14 CFR 125.127 - Location of fuel tanks.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Location of fuel tanks. 125.127 Section 125... Requirements § 125.127 Location of fuel tanks. (a) Fuel tanks must be located in accordance with § 125.153. (b... compartment may be used as the wall of an integral tank. (c) Fuel tanks must be isolated from personnel...

Monte Carlo modeling for realizing optimized management of failed fuel replacement

International Nuclear Information System (INIS)

Morishita, Kazunori; Yamamoto, Yasunori; Nakasuji, Toshiki

2014-01-01

Fuel cladding is one of the key components in a fission reactor to keep confining radioactive materials inside a fuel tube. During reactor operation, the cladding is however sometimes breached and radioactive materials leak from the fuel ceramic pellet into the coolant water through the breach. The primary coolant water is therefore monitored so that any leak is quickly detected, where the coolant water is periodically sampled and the concentration of, for example the radioactive iodine 131 (I-131), is measured. Depending on the measured concentration, the faulty fuel assembly with leaking rod is removed from the reactor and replaced by new one immediately or at the next refueling. In the present study, an effort has been made to develop a methodology to optimize the management for replacement of failed fuels due to cladding failures using the I-131 concentration measured in the sampled coolant water. A model numerical equation is proposed to describe the time evolution of I-131 concentration due to fuel leaks, and is then solved using the Monte-Carlo method as a function of sampling rate. Our results have indicated that, in order to achieve the rationalized management of failed fuels, higher resolution to detect a small amount of I-131 is not necessarily required but more frequent sampling is favorable. (author)

A combined wet/dry sipping cell for investigating failed triga fuel elements

International Nuclear Information System (INIS)

Boeck, H.; Gallhammer, H.; Hammer, J.; Israr, M.

1987-08-01

A sipping cell to detect failed triga fuel has been designed and constructed at the Atominstitut. The cell allows both wet- and dry sipping of one single standard triga fuel element. In the dry sipping method the fuel element may be electrically heated up to a maximum temperature of about 300 0 C to allow the detection of temperature dependent fission product release from the fuel element. 20 figs., 1 tab. (Author)

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

Methodology of failed section location in case of leaks in PGN-200M sectional steam generator

International Nuclear Information System (INIS)

Kuznetsov, A.A.; Govorov, P.P.; Nosov, Yu.V.; Karavaev, A.P.

2009-01-01

The article considers a way of the failed section location when indications of water-sodium reaction emerge in PGN-200M sectional steam generator of the BN-600 power unit. The selection of diagnostic parameters used to locate the failed section is justified. Various alternative locations of leaks have been simulated [ru

14 CFR 121.229 - Location of fuel tanks.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Location of fuel tanks. 121.229 Section 121... of fuel tanks. (a) Fuel tanks must be located in accordance with § 121.255. (b) No part of the engine... the wall of an integral tank. (c) Fuel tanks must be isolated from personnel compartments by means of...

Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

2006-01-01

The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts

International Nuclear Information System (INIS)

Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

2014-01-01

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of "1"3"4Cs and "1"3"7Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during the AGR-1 irradiation test or post-irradiation safety testing at 1600– 1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the ORNL Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking. (author)

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D., E-mail: hunnjd@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-09-15

Highlights: • Cesium release was used to detect SiC failure in HTGR fuel. • Tristructural-isotropic particles with SiC failure were isolated by gamma screening. • SiC failure was studied by X-ray tomography and SEM. • SiC degradation was observed after irradiation and subsequent safety testing. - Abstract: As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of {sup 134}Cs and {sup 137}Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were

Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

Apparatus for locating defective nuclear fuel elements

International Nuclear Information System (INIS)

Lawrie, W.E.

1979-01-01

An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

Method and apparatus for detecting failed fuels in nuclear reactor

International Nuclear Information System (INIS)

Tsuji, Tadashi.

1981-01-01

Purpose: To enable safety and automatic sampling for sample water in failed fuel detection. Constitution: A cap containing inner caps by the number of fuel assemblies inserted into each grid of a nuclear reactor is mounted to the upper end of the fuel assemblies. After the mounting, it is confirmed if the mounting is collectly made by the mounting state detection device utilizing the change in the pressure within the tube communicating to a water seal pipe. Then, air at a predetermined pressure introduced from an air supply tube opening into the cap is introduced into the cap to replace the coolants in the cap with the air. The pressure difference between the inside and the outside of the cap is detected and, if it shows a set value, it is confirmed that the cooling water level is independent for every fuel assembly. Then, sample water is sampled by the sampling tube within the guide cap provided to the upper end of the inner cap. (Furukawa, Y.)

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)

Failed fuel identification techniques for liquid-metal cooled reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

1995-01-01

The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

Failed (leaking) spent fuel management and storage in the Paks NPP

International Nuclear Information System (INIS)

Burjan, T.

2011-01-01

At the cycle 22, unit 4, Paks NPP the fissile contents raised irregularly in the water of the primary circuit. At the end of the cycle sipping tests were performed for the entire core to find out the leaking fuel assembly primarily responsible for this phenomenon. The identified leaking assembly temporarily was placed in the Spent Fuel Relaxing Pool. For measuring environmental impact of leaking assemblies an investigation program was developed and implemented. The assessment covered the following: effects of the leaking fuel on the water of relaxing pool and on the gaseous emissions in case open storage; in case when the leaking cassette is in a special hermetical storage case, how much gas is collected in the locked case and what is its composition; how to change the measured sipping test signal depending on relaxing time of leaking fuel cassettes. Based on the evaluation of the investigation program results the NPP modified the operational instructions for the treatment and storage of failed fuel assemblies. (author)

Detection and location of leaking TRIGA fuel elements

International Nuclear Information System (INIS)

Bouchey, G.D.; Gage, S.J.

1970-01-01

Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

An integrated approach for investigation of failed nuclear fuel used at NPP Cernavoda Unit 1

International Nuclear Information System (INIS)

Tuturici, I.L.; Parvan, M.; Popov, M.; Dobrin, R.; Staicu, C.

1996-01-01

At NPP Cernavoda-Unit 1 the fuel surveillance and the defect detection system in operation are based on monitoring the coolant activity concentration and on measuring the flux of delayed neutrons emitted by some short-lived fission products. In order to identify the failed fuel underwater non-destructive examination has to be performed. The major interest for the availability of underwater examination consists in the necessity of a speedy acquisition of the data on failed fuel in operation and of appropriate follow-up actions to be taken. Often the identification operation will be followed by more detailed examinations on selected fuel rods in the hot cells of the Post-irradiation Examination Laboratory of the Institute for Nuclear Research at Pitesti. Transfer of selected fuel rods will be done by the use of a type B(U) road transportation cask. Such an integrated approach will help to keep the level of activity concentration of the primary circuit well below the authorized limits. (author). 2 figs., 1 tab., 2 refs

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

Convective-diffusive transport of fission products in the gap of a failed fuel element

International Nuclear Information System (INIS)

Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

1995-03-01

A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

A follow-up test of failed fuel element of a nuclear reactor

International Nuclear Information System (INIS)

Peerasathien, W.

1974-01-01

This thesis is a result of test of a number of nuclear fuel rods which have not been used for a long time due to leakage of radioactivity. Water was circulated through each fuel rod in a test cylinder and radioactivity in water was measured. It was found that the detection of Cesium-137 which has a long half-life, does not indicate the extent of leakage of short-lived radioisotopes, some of which are gaseous. These gases are harmful to the reactor operators and users. A better result was obtained by placing the failed fuel rod in the test cylinder close to the reactor to induce fission. Short half-life gases or other nuclides of the same series were then directly measured

Pressurised light water reactor failed fuel monitoring. Comparison of international strategies used in the surveillance of the primary barrier

International Nuclear Information System (INIS)

Aldworth, Robin; Dinse, Christelle; Mole, Didier

2014-01-01

As part of its focus on continuous improvement, the nuclear industry has devoted considerable effort to the prevention, monitoring, characterisation and impact mitigation of fuel failures. Failed fuel and primary circuit activity monitoring strategies have thus evolved over time, with utilities changing their strategies as a function of operational experience, regulator encouragement and improvements to scientific understanding. Differences among utility fuel monitoring strategies are therefore to be expected. This paper compares and analyses a selection of failed fuel and primary circuit activity monitoring strategies employed at French, American, South-African and Belgian pressurised water reactors (PWRs). These strategies, which were received in response to a questionnaire sent by Electricité de France, are discussed with a number of potential best practices and areas for future research and development being provided. Finally, and so as to ensure the utility of this paper with regards to the development, or evolution of, failed fuel monitoring strategies, this paper has been split into four sections representing the primary radiochemistry indicator groups: gross, calculated, fission products and activated corrosion products. (author)

Studies on modeling to failed fuel detection system response in LMFBR

International Nuclear Information System (INIS)

Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

1981-05-01

Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Failed fuel rod detection system and computerized manipulator during outages

International Nuclear Information System (INIS)

Boehm, H.H.; Foerch, H.

1984-01-01

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

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

OpenAIRE

Asnavandi, Majid; Kahram, Mohaddeseh; Rezaei, Milad; Rezakhani, Davar

2017-01-01

The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tu...

Fuel failure detection and location methods in CAGRs

International Nuclear Information System (INIS)

Harris, A.M.

1982-06-01

The release of fission products from AGR fuel failures and the way in which the signals from such failures must be detected against the background signal from uranium contamination of the fuel is considered. Theoretical assessments of failure detection are used to show the limitations of the existing Electrostatic Wire Precipitator Burst Can Detection system (BCD) and how its operating parameters can be optimised. Two promising alternative methods, the 'split count' technique and the use of iodine measurements, are described. The results of a detailed study of the mechanical and electronic performance of the present BCD trolleys are given. The limited experience of detection and location of two fuel failures in CAGR using conventional and alternative methods is reviewed. The larger failure was detected and located using the conventional BCD equipment with a high confidence level. It is shown that smaller failures may not be easy to detect and locate using the current BCD equipment, and the second smaller failure probably remained in the reactor for about a year before it was discharged. The split count technique used with modified BCD equipment was able to detect the smaller failure after careful inspection of the data. (author)

Fail-safe system for activity cooled supersonic and hypersonic aircraft. [using liquid hydrogen fuel

Science.gov (United States)

Jones, R. A.; Braswell, D. O.; Richie, C. B.

1975-01-01

A fail-safe-system concept was studied as an alternative to a redundant active cooling system for supersonic and hypersonic aircraft which use the heat sink of liquid-hydrogen fuel for cooling the aircraft structure. This concept consists of an abort maneuver by the aircraft and a passive thermal protection system (TPS) for the aircraft skin. The abort manuever provides a low-heat-load descent from normal cruise speed to a lower speed at which cooling is unnecessary, and the passive TPS allows the aircraft skin to absorb the abort heat load without exceeding critical skin temperature. On the basis of results obtained, it appears that this fail-safe-system concept warrants further consideration, inasmuch as a fail-safe system could possibly replace a redundant active cooling system with no increase in weight and would offer other potential advantages.

Effect of Gas Fueling Location on H-mode Access in NSTX

International Nuclear Information System (INIS)

Maingi, R.; Bell, M.; Bell, R.; Biewer, T.; Bush, C.; Chang, C.S.; Gates, D.; Kaye, S.; Kugel, H.; LeBlanc, B.; Maqueda, R.; Menard, J.; Mueller, D.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

2003-01-01

The dependence of H-mode access on the poloidal location of the gas injection source has been investigated in the National Spherical Torus Experiment (NSTX). We find that gas fueling from the center stack midplane area produces the most reproducible H-mode access with generally the lowest L-H threshold power in lower single-null configuration. The edge toroidal rotation velocity is largest (in direction of the plasma current) just before the L-H transition with center stack midplane fueling, and then reverses direction after the L-H transition. Simulation of these results with a 2-D guiding-center Monte Carlo neoclassical transport code is qualitatively consistent with the trends in the measured velocities. Double-null discharges exhibit H-mode access with gas fueling from either the center stack midplane or center stack top locations, indicating a reduced sensitivity of H-mode access on fueling location in that shape

Aims of failed fuel detection and substantiation of radiation safety at implementation of new kinds of nuclear fuel and fuel cycles on NPP with WWER

International Nuclear Information System (INIS)

Miglo, V.; Luzanova, L.

2011-01-01

Limiting of number of leaking fuel rods in a core during reactor operation in the analyses which are carried out for a substantiation of radiating safety for NPP with WWER as well as problems and possibilities of FFD at implementation of new kinds of fuel and fuel cycles are the main topics discussed in this paper. Available experience of designing of the NPP with WWER shows, that for ensuring of implementation of the RS criteria regarding limiting radioactive emissions from the NPP and doses of an irradiation of the population living near to NPP, it is required to regulate more rigidly number of failed fuel rods in comparison with requirements of Rules of nuclear safety NP-082-07. The reason of it is necessity to consider a technical condition of all safety barriers on a path of radioactive FP extension in a complex, first and foremost of uncontrolled leakage of the primary coolant to the NPP premises and efficiency of filters of ventilating systems, and also spike-effect on activity of isotopes of iodine after a power unit shutdown for fuel reloading and openings of a cover of a reactor. Depending on the project of NPP, parameters of fuel loading, a place of placing of the NPP and other factors the limit level of activity of isotopes of iodine in the primary coolant will be reached at various number of leaking fuel rods which can be unequal for various power units and the NPP with WWER, constructed on one design. The quantity of leaking fuel rods at which the design limit on FP-activity in the primary coolant of operating reactor is reached, can be essential below an operational limit on number of failed fuel rods established by Rules of nuclear safety. However the reached quality of fabrication of the WWER fuel rods providing their high reliability (the probability of fuel rod failure in the course of one operation year is not higher than 10 -5 ) as well as due to the levels of the WWER fuel rod depressurization actually attainable in the normal conditions of

Studying some regimes of the WWER-440 type reactor failed fuel element operation

International Nuclear Information System (INIS)

Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

1981-01-01

The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

Fundamental principles of failed fuel detection concepts on nuclear power units of WWER type

International Nuclear Information System (INIS)

Lusanova, L.; Miglo, V.; Slavyagin, P.

2001-01-01

The subject of the paper is the Russian failed fuel detection concept in both operating and shut down reactors. The philosophy for detection of fission products released from defective fuel during operation and sipping tests and using of these results for regulation of the radiological situation at the NPP during the next cycle is widely spread. In presented work such philosophy is applied to the shut down rectors. An option for sipping test performed in a mast of Refueling Machine (RM) using a wet-gas version of sipping test is briefly described. The use of the FFD method in RM mast allows combining the procedure of Fuel Assemblies (FA) tightness test with transport operation during reloading of the fuel from the core into the cooling pool. This reduces the time for reloading and transport operation with FA and increases the safety of reactor operation. The FFD method in RM mast has passed successful tests on Unit 4 at Balakovskaja NPP and it is expected to apply in other NPP unit with WWER-1000 reactors

Method of detecting failed fuels for a reactor

International Nuclear Information System (INIS)

Urata, Megumu; Kamiya, Kunio.

1982-01-01

Purpose: To detect the failed fuel with high sensitivity without causing insensitive measuring zone and measuring efficiency decrease time zone even if the temperature and the flow rate of sampling gas vary. Constitution: The simultaneous ventilation times t 2 of activated charcoal adsorption towers A and B are set to a value less than the holding-up time of primary radioactive nuclide in the one tower A, more than the time reaching the ordinary flowout state of secondary radioactive nuclide in the other tower B and yet less than 1/2 of the remaining time t 6 obtained by subtracting the switching time t 1 of the one tower A by the retreating time t 5 of the adsorbent. In the case that the ventilation to the one tower A and the measurement of the radioactivity level is switched to the other tower after the simultaneous ventilation in this manner, the flowout of the secondary radioactive nuclide in the other tower B becomes the ordinary state. Accordingly, no measuring efficiency decreasing time zone and insensitive measurement zone will not take place. (Seki, T.)

Failed fuel monitoring at nuclear power plants with RBMK reactors: operating parameters, requirements and decision making criteria

International Nuclear Information System (INIS)

Zhukov, I.V.

1993-01-01

The procedure for estimating the number of failed fuel rods in the core and the prediction of their discharge efficiency during operation is presented. The procedure is based on the FFM data base and the I-131 and Xe-133 coolant activity. (author)

75 FR 59673 - Public Hearing Locations for the Proposed Fuel Economy Labels

Science.gov (United States)

2010-09-28

...] RIN 2060-AQ09; RIN 2127-AK73 Public Hearing Locations for the Proposed Fuel Economy Labels AGENCY... Vehicle Fuel Economy Label,'' published in the Federal Register on September 23, 2010. The goal of a... testimony or comment on the Agency's proposed revisions and additions to the motor vehicle fuel economy...

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

Contamination of a PWR primary circuit by fuel pins with failed cladding

International Nuclear Information System (INIS)

Janvier, J.C.; Chagrot, M.

1979-01-01

The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

Energy Technology Data Exchange (ETDEWEB)

Strindehag, O

1966-06-15

Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type ({nu} = 10{sup -8}/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the {gamma}-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A{sup 41} in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems.

Optimum MRS site location to minimize spent fuel transportation impacts

International Nuclear Information System (INIS)

Hoskins, R.E.

1987-01-01

A range of spent fuel transportation system parameters are examined in terms of attributes important to minimizing transportation impacts as a basis for identifying geographic regions best suited for siting a monitored retrievable storage (MRS) facility. Transportation system parameters within existing transport cask design and transportation mode capabilities were systematically analyzed. The optimum MRS location was found to be very sensitive to transportation system assumptions particularly with regard to the relative efficiencies of the reactor-to-MRS and MRS-to-repository components of the system. Moreover, dramatic improvements in the reactor-to-MRS component can be made through use of multiple cask shipment of the largest practical casks by dedicated train compared to the traditional single cask rail (70%) and truck (30%) shipments assumed the Department of Energy in their studies that defined the optimum MRS location in the vicinity of Tennessee. It is important to develop and utilize an efficient transportation system irrespective of whether or not an MRS is in the system. Assuming reasonably achievable efficiency in reactor-to-MRS spent fuel transportation and assigning equal probabilities to the three western sites selected for characterization of being the repository site, the optimum MRS location would be in the far-mid-western states. Based on various geographic criteria including barge access and location in a nuclear service area, the State of Tennessee ranks any place from 12th to the 25th at a penalty of about 30% over the minimum achievable impacts. While minimizing transportation impacts is an important factor, other criteria should also be considered in selecting an MRS site

A Comparison of Fueling with Deuterium Pellet Injection from Different Locations on the DIII-D Tokamak

International Nuclear Information System (INIS)

Baylor, L.R.; Combs, S.K.; Gohil, P.; Houlberg, W.A.; Hsieh, C.; Jernigan, T.C.; Parks, P.B.

1999-01-01

Initial pellet injection experiments on DIII-D with high field side (HFS) injection have demonstrated that deeper pellet fuel deposition is possible even with HFS injected pellets that are significantly slower than pellets injected from the low field side (LFS) (outer midplane) location. A radial displacement of the pellet mass shortly after or during the ablation process is consistent with the observed mass deposition profiles measured shortly after injection. Vertical injection inside the magnetic axis shows some improvement in fueling efficiency over LFS injection and may provide an optimal injection location for fueling with high speed pellets

A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL & TREATMENT & DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

Energy Technology Data Exchange (ETDEWEB)

PERES, M.W.

2007-01-09

Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel.

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

Directory of Open Access Journals (Sweden)

Majid Asnavandi

2017-01-01

Full Text Available The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tubes is not uniform which suggests that fire-side corrosion has happened on the tubes. Fire-side corrosion is caused by the reaction of combustion products with oxide layers on the tube surface resulting in metal loss and consequently tubes fracture. However, the tubes corrosion behaviour did not follow the conventional models of the fire-side corrosion. Given that, using the corrosion monitoring techniques for these boiler tubes seems essential. As a result, the thickness of the boiler tubes in different parts of the boiler has been recorded and critical points are selected accordingly. Such critical points are selected for installation of corrosion monitoring probes.

Comparison of two inductive learning methods: A case study in failed fuel identification

Energy Technology Data Exchange (ETDEWEB)

Reifman, J. [Argonne National Lab., IL (United States); Lee, J.C. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering

1992-05-01

Two inductive learning methods, the ID3 and Rg algorithms, are studied as a means for systematically and automatically constructing the knowledge base of expert systems. Both inductive learning methods are general-purpose and use information entropy as a discriminatory measure in order to group objects of a common class. ID3 constructs a knowledge base by building decision trees that discriminate objects of a data set as a function of their class. Rg constructs a knowledge base by grouping objects of the same class into patterns or clusters. The two inductive methods are applied to the construction of a knowledge base for failed fuel identification in the Experimental Breeder Reactor II. Through analysis of the knowledge bases generated, the ID3 and Rg algorithms are compared for their knowledge representation, data overfitting, feature space partition, feature selection, and search procedure.

Comparison of two inductive learning methods: A case study in failed fuel identification

Energy Technology Data Exchange (ETDEWEB)

Reifman, J. (Argonne National Lab., IL (United States)); Lee, J.C. (Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering)

1992-01-01

Two inductive learning methods, the ID3 and Rg algorithms, are studied as a means for systematically and automatically constructing the knowledge base of expert systems. Both inductive learning methods are general-purpose and use information entropy as a discriminatory measure in order to group objects of a common class. ID3 constructs a knowledge base by building decision trees that discriminate objects of a data set as a function of their class. Rg constructs a knowledge base by grouping objects of the same class into patterns or clusters. The two inductive methods are applied to the construction of a knowledge base for failed fuel identification in the Experimental Breeder Reactor II. Through analysis of the knowledge bases generated, the ID3 and Rg algorithms are compared for their knowledge representation, data overfitting, feature space partition, feature selection, and search procedure.

Comparison of two inductive learning methods: A case study in failed fuel identification

International Nuclear Information System (INIS)

Reifman, J.; Lee, J.C.

1992-01-01

Two inductive learning methods, the ID3 and Rg algorithms, are studied as a means for systematically and automatically constructing the knowledge base of expert systems. Both inductive learning methods are general-purpose and use information entropy as a discriminatory measure in order to group objects of a common class. ID3 constructs a knowledge base by building decision trees that discriminate objects of a data set as a function of their class. Rg constructs a knowledge base by grouping objects of the same class into patterns or clusters. The two inductive methods are applied to the construction of a knowledge base for failed fuel identification in the Experimental Breeder Reactor II. Through analysis of the knowledge bases generated, the ID3 and Rg algorithms are compared for their knowledge representation, data overfitting, feature space partition, feature selection, and search procedure

Failed fuel detection device

International Nuclear Information System (INIS)

Kasahara, Yoshiyuki; Soroi, Masatoshi.

1992-01-01

A pair of coil springs each of different spring rigidity are disposed independently to an interface mechanism which engages a reactor core fuel assembly. The springing reaction of the coil springs is utilized for providing a structure capable of detaching. A driving portion vertically movable in an inner cylinder of a system main body interlocking with the intrface mechanism is disposed, as well as a system separation mechanism is disposed for conducting electromotive remote control when it is required. With such a constitution, although it has been necessary so far that a plurality of operators access the reactor core upper mechanisms, it is not necessary according to the device of the present invention. Accordingly, the problem of operator's exposure can be overcome. (I.S.)

Failed fuel rod detection method by ultrasonic wave

International Nuclear Information System (INIS)

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

1990-01-01

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

Repairing fuel for reinsertion

International Nuclear Information System (INIS)

Krukshenk, A.

1986-01-01

Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL, TREATMENT and DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

International Nuclear Information System (INIS)

PERES, M.W.

2007-01-01

Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel

Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

International Nuclear Information System (INIS)

Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

2009-01-01

The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.

Effect of spark plug and fuel injector location on mixture stratification in a GDI engine - A CFD analysis

Science.gov (United States)

Saw, O. P.; Mallikarjuna, J. M.

2017-09-01

The mixture preparation in gasoline direct injection (GDI) engines operating at stratified condition plays an important role in deciding the combustion, performance and emission characteristics of the engine. In a wall-guided GDI engine, with a late fuel injection strategy, piston top surface is designed in such a way that the injected fuel is directed towards the spark plug to form a combustible mixture at the time of ignition. In addition, in these engines, location of spark-plug and fuel injector, fuel injection pressure and timing are also important to create a combustible mixture near the spark plug. Therefore, understanding the mixture formation under the influence of the location of spark plug and fuel injector is very essential for the optimization of the engine parameters. In this study, an attempt is made to understand the effect of spark plug and fuel injector location on the mixture preparation in a four-stroke, four-valve and wall-guided GDI engine operating under a stratified condition by using computational fluid dynamics (CFD) analysis. All the CFD simulations are carried out at an engine speed of 2000 rev/min., and compression ratio of 10.6, at an overall equivalence ratio (ER) of about 0.65. The fuel injection and spark timings are maintained at 605 and 710 CADs respectively. Finally, it is concluded that, combination of central spark plug and side fuel injector results in better combustion and performance.

Liquid water breakthrough location distances on a gas diffusion layer of polymer electrolyte membrane fuel cells

Science.gov (United States)

Yu, Junliang; Froning, Dieter; Reimer, Uwe; Lehnert, Werner

2018-06-01

The lattice Boltzmann method is adopted to simulate the three dimensional dynamic process of liquid water breaking through the gas diffusion layer (GDL) in the polymer electrolyte membrane fuel cell. 22 micro-structures of Toray GDL are built based on a stochastic geometry model. It is found that more than one breakthrough locations are formed randomly on the GDL surface. Breakthrough location distance (BLD) are analyzed statistically in two ways. The distribution is evaluated statistically by the Lilliefors test. It is concluded that the BLD can be described by the normal distribution with certain statistic characteristics. Information of the shortest neighbor breakthrough location distance can be the input modeling setups on the cell-scale simulations in the field of fuel cell simulation.

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

International Nuclear Information System (INIS)

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

1978-05-01

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

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1990-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous radionuclides. A modeling approach is outlined and sample calculations are provided that show the effects on release due to a limited perforation area, to decreasing temperature, and to the partial occlusion of the perforated area by corrosion products. 8 refs., 2 tabs

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1989-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous 14 CO 2 . A modeling approach is outlined and sample calculations are provided that show the effect on release due to a limited perforation area, to decreasing temperature, and to partial occlusion of the perforated area by corrosion products. 5 refs., 4 tabs

Fuel Cell Demonstration Project - 200 kW - Phosphoric Acid Fuel Cell Power Plant Located at the National Transportation Research Center: FINAL REPORT

Energy Technology Data Exchange (ETDEWEB)

Berry, JB

2005-05-06

Oak Ridge National Laboratory (ORNL) researches and develops distributed generation technology for the Department of Energy, Energy Efficiency and Renewable Energy Distributed Energy Program. This report describes installation and operation of one such distributed generation system, a United Technology Corporation fuel cell located at the National Transportation Research Center in Knoxville, Tennessee. Data collected from June 2003 to June of 2004, provides valuable insight regarding fuel cell-grid compatibility and the cost-benefit of the fuel cell operation. The NTRC fuel cell included a high-heat recovery option so that use of thermal energy improves project economics and improves system efficiency to 59% year round. During the year the fuel cell supplied a total of 834MWh to the NTRC and provided 300MBtu of hot water. Installation of the NTRC fuel cell was funded by the Distributed Energy Program with partial funding from the Department of Defense's Climate Change Fuel Cell Buy Down Program, administered by the National Energy Technology Laboratory. On-going operational expenses are funded by ORNL's utility budget and are paid from operational cost savings. Technical information and the benefit-cost of the fuel cell are both evaluated in this report and sister reports.

A study on dissolution and leaching behaviour of spent nuclear fuels

International Nuclear Information System (INIS)

Lee, Chang Heon; Im, Hee Jung; Kim, Jong Gu; Park, Yang Soon; Ha, Yeong Keong

2010-12-01

This state of the art report describes a leaching behaviour of spent nuclear fuels which should be considered for safety assessment of spent nuclear fuel disposal in a deep geological repository. A decisive factor of a dissolution of UO 2 , a matrix of the fuel, is chemical characters (redox potential, pH, concentration of inorganic anions, water radiolysis subsequent by radiation field of the fuels) of ground water expected to be in contact with the fuels after the container has failed due to corrosion as well as atmosphere condition of a deep geological repository, which can change the oxidation state of UO 2 . The release rates of radionuclides from UO 2 matrix depend largely on their location within the fuels, that is, the radionuclides fixed in the fuel/cladding gap and grain boundaries are rapidly released. However, the radionuclides within the grains of the fuel are slowly released, and then their release rate is governed by a dissolution behaviour of UO 2

Placed Parenting, locating unrest: failed femininities, troubled mothers and rioting subjects

Directory of Open Access Journals (Sweden)

Kim Allen

2012-07-01

Full Text Available Critical attention has been given to the consolidation of classed forms of 'placed personhood', as compelling future-orientated and self-regulating subjects that 'fit' into contemporary economic and social formation. These forms of personhood, spoken of as moral character and behavioural 'traits', are increasingly attached to placed parenthood: as that which (selflocates in the right moral and material terrain. Good subjects – made through good parenting and in particular via 'good mothers' – are tasked with self-optimizing and bringing forward their own futures and those of their families (Allen and Osgood 2009; Armstrong 2010; Evans 2010, Gillies 2007; Lawler 2000; Taylor 2012a; Taylor and Addison 2011. A broader 'public' as concerned with equity, welfare and redistribution is dis-placed and entirely re-placed with a self-orientated 'enterprising' privatized response. This limited response is self-congratulating of its own 'responsibility' and condemning of those who 'fail' heightened efforts (in times of 'cutting back' parent-citizens are told simply to be more 'efficient'. 'Necessity' and 'austerity' are, as the editors of this special issue highlight, invoked to re-do all kind of classed and gendered violence, where the most privileged sections of society are evacuated from blame in times of economic crisis: in contrast, what re-circulates is a stated need for poor, 'failing mothers' to 'step-up' for all our sakes. Those who cannot bring themselves forward and propel into this neo-liberal future are increasingly condemned as the wrong kind of parents. This piece aims to chart some of the 'ugliness of parenting' (Taylor 2009, 2012b as placed parenthood attaches to specific classed locales and subjects as that which 'fails' and 'troubles' the future. This is witnessed in the example of and responses to the English Riots of 2011, which is placed as a case study casting light on intersections of class, race, gender and sexuality in forms

Improvement of failed fuel detection system of light water reactor

International Nuclear Information System (INIS)

Chung, M.K.; Kang, H.D.; Cho, S.W.; Lee, K.W.

1981-01-01

Multi-task DAAS system by utilizing PDP-11/23 computer was assembled and tested for its performances. By connecting four Ge(Li) detectors to this DAAS, test experiments were done to prove system capability for detection and analysis of any fission gases resolved in four independently sampled primary cooling water from a power reactor. Appropriate computer programs were also introduced for this application and satisfactory results were obtained. Further application of this DAAS to the quality test of fuel pins (uniform distribution of enriched uranium in fresh fuel pellets), a prototype fuel scanner system was designed, constructed and tested. Operational principle of this system is based on the determination of 235 U/ 238 U abundance ratio in pellets by precision spectrometry or gamma-rays which are emitted from a portion of fuel pellets. For the uniform scanning, rotational and traverse motions at pre-selected speeds were applied to a fuel pin under tests. A long lens magnetic beta-spectrometer of Argonne National Laboratory was transferred to KAERI and re-installed for future precision beta-gamma spectroscopic research works on short-lived fission products nuclei

Fuel failure assessments based on radiochemistry. Experience feedback and challenges

International Nuclear Information System (INIS)

Petit, C.; Ziabletsev, D.; Zeh, P.

2015-01-01

Significant improvements have been observed in LWR nuclear fuel reliability over the past years. As a result, the number of fuel failures in PWRs and BWRs has recently dramatically decreased. Nevertheless, a few remaining challenges still exist. One of them is that the industry has recently started seeing a relatively new type of fuel failure, so-called 'weak leak failures', which could be characterized by a very small release of gaseous fission products and essentially almost zero release of iodines or any other soluble fission products in the reactor coolant. Correspondingly, the behavior of these weak leakers does not follow typical behavior of a conventional leaker characterized by a proportionality of the amount of released Xe 133 related to the failed rod power. Instead, for a weak leaker, the activity of Xe 133 is directly correlated to the size of the cladding defects of the leaker. The presence of undetected weak leaker in the core may lead to carryover of a leaker into the subsequent cycle. Even if the presence of weak leaker in the core is suspected, it typically requires more effort to identify the leaker which could result in extended duration of the outage and ultimately to economic losses to the utility operating the reactor. To effectively deal with this issue the industry has been facing, several changes have been recently realized, which are different from the methodology of dealing with conventional leaker. These changes include new assessment methods, the need for improved sipping techniques to better identify low release leakers, and correspondingly better equipment to be able to locate small clad defects associated with weak leaker, such as sensitive localization device of failed rods, sensitive eddy current coil for the failed rod, ultra high definition cameras for the failed rod examination and experienced fuel reliability engineers performing cause of failure and rood cause research and analyses. Ultimately, the destructive

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

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1992-03-01

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

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

Gas release from a failed fuel pin after reactor shut-down

International Nuclear Information System (INIS)

Pshenichnikov, B.V.

1975-01-01

A mathematical model of gassing from a hypothetical core fuel element in the active zone of a stopped water-moderated reactor was analysed to investigate the process of liberation of gaseous fission products from an unpressurized fuel element. A one-dimensional problem was obtained as a result of the accepted hypotheses. A fault was assumed to have occured during reactor operation; at the same time, a vapour-gas mixture was considered to be present under the envelope at reactor working pressure by the moment of stoppage. An approximative estimation was made of the retardation time of pressure balancing at the open end of the fuel element, and also of the amount of total gas remaining in the gap under the fuel element envelope after pressure drop in the reactor. Estimation of retardation time permitted to conclude that pressure in the nonhermetic fuel element envelope follows pressure fluctuation in the reactor in the course of cooling, the retardation time of pressure balancing outside and inside the fuel element lasting but a few seconds

Development of treatment technology of radio-contaminated coolant in fuel test loop

International Nuclear Information System (INIS)

Kim, J. Y.

1997-10-01

In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs

Development of treatment technology of radio-contaminated coolant in fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-10-01

In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs.

Effects of thermocouple installation and location on fuel rod temperature measurements

International Nuclear Information System (INIS)

McCormick, R.D.

1983-01-01

This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

Alternative Fuels Data Center: Biodiesel Fueling Stations

Science.gov (United States)

Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Stations Photo of a location or along a route. Infrastructure Development Learn about biodiesel fueling infrastructure codes Case Studies California Ramps Up Biofuels Infrastructure Green Fueling Station Powers Fleets in Upstate

Alternative Fuels Data Center: Ethanol Fueling Stations

Science.gov (United States)

... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

Plan of studies on fuel failure detection in Rikkyo Research Reactor

International Nuclear Information System (INIS)

Matsuura, T.; Nagahara, T.; Hattori, M.; Kawaguchi, K.

1987-01-01

Studies on fuel failure detection in Rikkyo Research Reactor have recently been begun in the following four approaches. (1) Accumulation of the data on the concentration of the short-lived radioactivity originating from FP rare gases contained in the air on the water surface of the reactor tank. (2) Accumulation of the data on the concentration of FP (especially 131 I) in the water of the reactor tank. (3) Design and preparation of a ''sniffer'' by which the location of the failed fuel element can be detected, when some anomaly is found in the above two routine measurements. (4) Design and preparation of a vessel containing a fuel element, which can be useful both for ''sipping'' inspection of the fuel element and for storage of the damaged fuel element. In this paper, an outline of the above approaches and the results of some preliminary experiments are reported. (author)

KNF's fuel service technologies and experiences

International Nuclear Information System (INIS)

Shin, Jung Cheol; Kwon, Jung Tack; Kim, Jaeik; Park, Jong Youl; Kim, Yong Chan

2009-01-01

In Korea, since 1978, the commercial nuclear power plant was operated. After 10 years, from 1988, the nuclear fuel was produced by KNF (Korea Nuclear Fuel). The Fuel Service Team was established at KNF in 1995. Through the technical self reliance periods in cooperate with advanced foreign companies for 5 years, KNF has started to carry out fuel service activities onsite in domestic nuclear power plants. By ceaseless improving and advancing our own methodologies, after that, KNF is able to provide the most safe and reliable fuel repair services and poolside examinations including the root cause analysis of failed fuels. Recently, KNF developed the fuel cleaning system using ultrasonic technique for crud removal, and the CANDU fuel sipping system to detect a failed fuel bundle in PHWR. In this paper, all of KNF's fuel service technologies are briefly described, and the gained experience in shown

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.

Conditions inside Water Pooled in a Failed Nuclear Waste Container and its Effect on Radionuclide Release

Science.gov (United States)

Hamdan, L. K.; Walton, J. C.; Woocay, A.

2009-12-01

Nuclear power use is expected to expand in the future, as part of the global clean energy initiative, to meet the world’s surging energy demand, and attenuate greenhouse gas emissions, which are mainly caused by fossil fuels. As a result, it is estimated that hundreds of thousands of metric tons of spent nuclear fuel (SNF) will accumulate. SNF disposal has major environmental (radiation exposure) and security (nuclear proliferation) concerns. Storage in unsaturated zone geological repositories is a reasonable solution for dealing with SNF. One of the key factors that determine the performance of the geological repository is the release of radionuclides from the engineered barrier system. Over time, the nuclear waste containers are expected to fail gradually due to general and localized corrosions and eventually infiltrating water will have access to the nuclear waste. Once radionuclides are released, they will be transported by water, and make their way to the accessible environment. Physical and chemical disturbances in the environment over the container will lead to different corrosion rates, causing different times and locations of penetration. One possible scenario for waste packages failure is the bathtub model, where penetrations occur on the top of the waste package and water pools inside it. In this paper the bathtub-type failed waste container is considered. We shed some light on chemical and physical processes that take place in the pooled water inside a partially failed waste container (bathtub category), and the effects of these processes on radionuclide release. Our study considers two possibilities: temperature stratification of the pooled water versus mixing process. Our calculations show that temperature stratification of the pooled water is expected when the waste package is half (or less) filled with water. On the other hand, when the waste package is fully filled (or above half) there will be mixing in the upper part of water. The effect of

Criteria for recladding of spent light water reactor fuel before long term pool storage

International Nuclear Information System (INIS)

Pettersson, K.; Jansson, L.

1979-01-01

The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that these degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods. (Auth.)

Criteria for recladding of spent light water reactor fuel before long term pool storage

International Nuclear Information System (INIS)

Pettersson, K.; Jansson, L.

1979-06-01

The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that thede degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods.(author)

An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

International Nuclear Information System (INIS)

Gillespie, G.E.

An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

Building a panel data set on fuel stations located in the Spanish regional areas of Madrid and Barcelona

Directory of Open Access Journals (Sweden)

Jacint Balaguer

2016-06-01

Full Text Available The data described in this article were collected daily over the period June 10, 2010, to November 25, 2012, from the website of the Spanish Ministry of Industry, Energy and Tourism. The database includes information about fuel stations regarding to their prices (both gross and net of taxes, brand, location (latitude and longitude, and postal code in the Spanish provinces of Madrid and Barcelona. Moreover, obtaining the postal codes has allowed us to select those stations that are operating within the metropolitan areas of Madrid and Barcelona. By considering those fuel stations that uninterruptedly provided prices during the entire period, the data can be especially useful to explore the dynamics of prices in fuel markets. This is the case of Balaguer and Ripollés (2016, “Asymmetric fuel price responses under heterogeneity” [1], who, taking into account the presence of the potential heterogeneity of the behaviour of fuel stations, used this statistical information to perform an analysis on asymmetric fuel price responses.

Prescribed burning in ponderosa pine: fuel reductions and redistributing fuels near boles to prevent injury

Science.gov (United States)

Prescribed burning can be an effective tool for thinning forests and reducing fuels to lessen wildfire risks. However, prescribed burning sometimes fails to substantially reduce fuels and sometimes damages/kills valuable, large trees. This study compared fuel reductions between fall and spring pre...

Failing Failed States

DEFF Research Database (Denmark)

Holm, Hans-Henrik

2002-01-01

coverage. A Danish survey of newsrooms shows that the national world-view and prevalent news criteria prevent consistent coverage. It is argued that politicians are the ones who determine national agendas: it is from political initiatives, rather than media coverage, that failing states and humanitarian......When states are failing, when basic state functions are no longer carried out, and when people have no security, humanitarian crises erupt. In confronting this problem, the stronger states have followed an ad hoc policy of intervention and aid. In some cases, humanitarian disasters have resulted...... from inaction. Often, the media are blamed. Politicians complain about the media when they interfere (the CNN effect), and when they do not. This article looks at how the media do cover failing states. Sierra Leone and Congo are used as examples. The analysis shows that there is little independent...

Fail-safe first wall for preclusion of little leakage

International Nuclear Information System (INIS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

Leakages although excluded by design measures would occur most probably in highly stressed areas, weldments and locations without possibility to classify the state by in-service inspection. In a water-cooled first wall, allowable leak rate of water is generally very small, and therefore, locating of the leak portion under highly activated environment will be very difficult and be time-consuming. The double-wall concept is promising for the ITER first wall, because it can be made fail-safe by the application of the leak-before-break and the multiple load path concepts, and because it has a potential capability to solve the little leak problem. When the fail safe strength is well defined, subcritical crack growth in the damaged wall can be permitted. This will enable to detect stable leakage of coolant without deteriorating plasma operation. The paper deals with the little leak problem and presents method for evaluating small leak rate of a liquid coolant from crack-like defects. The fail-safe first wall with the double-wall concept is also proposed for preclusion of little leakage and its fail-safety is discussed. (author)

Fuel failure monitoring system design approach for KALIMER

International Nuclear Information System (INIS)

Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

1998-01-01

Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

Long term fuel price elasticity: effects on mobility tool ownership and residential location choice - Final report

Energy Technology Data Exchange (ETDEWEB)

Erath, A.; Axhausen, K. W.

2010-04-15

This comprehensive final report for the Swiss Federal Office of Energy (SFOE) examines the long-term effects of fuel price elasticity. The study analyses how mobility tool usage and ownership as well as residence location choice are affected by rising fuel costs. Based on econometric models, long-term fuel price elasticity is derived. The authors quote that the demand reactions to higher fuel prices mainly observed are the reduction of mileage and the consideration of smaller-engined and diesel-driven cars. As cars with natural gas powered engines and electric drives were hardly considered in the survey, the results of the natural gas model can, according to the authors, only serve as a trend. No stable model could be estimated for the demand and usage of electric cars. A literature overview is presented and the design of the survey is discussed, whereby socio-demographical variables and the effects of price and residence changes are discussed. Modelling of mobility tool factors and results obtained are looked at. Finally, residence choice factors are modelled and discussed. Several appendices complete the report.

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

KNF's fuel service technologies and experiences

Energy Technology Data Exchange (ETDEWEB)

Shin, Jung Cheol; Kwon, Jung Tack; Kim, Jaeik; Park, Jong Youl; Kim, Yong Chan [KNF, Daejeon (Korea, Republic of)

2009-04-15

In Korea, since 1978, the commercial nuclear power plant was operated. After 10 years, from 1988, the nuclear fuel was produced by KNF (Korea Nuclear Fuel). The Fuel Service Team was established at KNF in 1995. Through the technical self reliance periods in cooperate with advanced foreign companies for 5 years, KNF has started to carry out fuel service activities onsite in domestic nuclear power plants. By ceaseless improving and advancing our own methodologies, after that, KNF is able to provide the most safe and reliable fuel repair services and poolside examinations including the root cause analysis of failed fuels. Recently, KNF developed the fuel cleaning system using ultrasonic technique for crud removal, and the CANDU fuel sipping system to detect a failed fuel bundle in PHWR. In this paper, all of KNF's fuel service technologies are briefly described, and the gained experience in shown.

A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

International Nuclear Information System (INIS)

Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

2001-05-01

In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

Reactor fuel exchanging facility

International Nuclear Information System (INIS)

Kubota, Shin-ichi.

1981-01-01

Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

Nuclear fuel shipping inspection device

International Nuclear Information System (INIS)

Takahashi, Toshio; Hada, Koji.

1988-01-01

Purpose: To provide an nuclear fuel shipping inspection device having a high detection sensitivity and capable of obtaining highly reliable inspection results. Constitution: The present invention concerns a device for distinguishing a fuel assembly having failed fuel rods in LMFBR type reactors. Coolants in a fuel assembly to be inspected are collected by a sampling pipeway and transferred to a filter device. In the filter device, granular radioactive corrosion products (CP) in the coolants are captured, to reduce the background. The coolants, after being passed through the filter device, are transferred to an FP catching device and gamma-rays of iodine and cesium nuclides are measured in FP radiation measuring device. Subsequently, the coolants transferred to a degasing device to separate rare gas FP in the coolants from the liquid phase. In a case if rare gas fission products are detected by the radiation detector, it means that there is a failed fuel rod in the fuel assembly to be inspected. Since the CP and the soluble FP are separated and extracted for the radioactivity measurement, the reliability can be improved. (Kamimura, M.)

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

Optimization of time and location dependent spent nuclear fuel storage capacity

International Nuclear Information System (INIS)

Macek, V.

1977-01-01

A linear spent fuel storage model is developed to identify cost-effective spent nuclear fuel storage strategies. The purpose of this model is to provide guidelines for the implementation of the optimal time-dependent spent fuel storage capacity expansion in view of the current economic and regulatory environment which has resulted in phase-out of the closed nuclear fuel cycle. Management alternatives of the spent fuel storage backlog, which is created by mismatch between spent fuel generation rate and spent fuel disposition capability, are represented by aggregate decision variables which describe the time dependent on-reactor-site and off-site spent fuel storage capacity additions, and the amount of spent fuel transferred to off-site storage facilities. Principal constraints of the model assure determination of cost optimal spent fuel storage expansion strategies, while spent fuel storage requirements are met at all times. A detailed physical and economic analysis of the essential components of the spent fuel storage problem, which precedes the model development, assures its realism. The effects of technological limitations on the on-site spent fuel storage expansion and timing of reinitiation of the spent fuel reprocessing on optimal spent fuel storage capacity expansion are investigated. The principal results of the study indicate that (a) expansion of storage capacity beyond that of currently planned facilities is necessary, and (b) economics of the post-reactor fuel cycle is extremely sensitive to the timing of reinitiation of spent fuel reprocessing. Postponement of reprocessing beyond mid-1982 may result in net negative economic liability of the back end of the nuclear fuel cycle

Some factors to consider in handling and storing spent fuel

International Nuclear Information System (INIS)

Bailey, W.J.

1985-11-01

This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

Improved fueling and transport barrier formation with pellet injection from different locations on DIII-D

International Nuclear Information System (INIS)

Baylor, L.R.; Jernigan, T.C.; Gohil, P.

2001-01-01

Pellet injection has been employed on DIII-D from different injection locations to optimize the mass deposition for density profile control and internal transport barrier formation. Transport barriers have been formed deep in the plasma core with central mass deposition from high field side (HFS) injected pellets and in the edge with pellets that trigger L-mode to H-mode transitions. Pellets injected from all locations can trigger the H-mode transition, which depends on the edge density gradient created and not on the radial extent of the pellet deposition. Pellets injected from inside the magnetic axis from the inner wall or vertical port lead to stronger central mass deposition than pellets injected from the low field side (LFS) and thus yield deeper more efficient fueling. (author)

Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

International Nuclear Information System (INIS)

Terasvirta, R.

2001-01-01

This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

Studies and research concerning BNFP: evaluation of spent-fuel-examination techniques for the Barnwell Nuclear Fuel Plant

International Nuclear Information System (INIS)

Anderson, R.T.; Gray, J.H.; Rogell, M.L.

1982-09-01

A study was made of various examinations which could be remotely performed on a production basis with spent fuel at the Barnwell Nuclear Fuel Plant (BNFP). These techniques could form an integral portion of fuel disassembly and canning operations. Their benefits accrue to either improved fuel storage, reprocessing, or both. In conjunctoin with these studies, evaluations have been made of the operational impact of receiving failed or canned fuel at the BNFP

Fuel element services

International Nuclear Information System (INIS)

Marta, H.; Alvarez, P.; Jimenez, J.

2006-01-01

Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

Shielding considerations for advanced fuel irradiation experiments

International Nuclear Information System (INIS)

Kang, Young-Hwan; Kim, Hee-Moon; Kim, Bong-Goo; Kim, Hark-Rho; Lee, Dong-Soo

2008-01-01

An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 μSv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels. (author)

TMI-2 spent fuel shipping

International Nuclear Information System (INIS)

Quinn, G.J.; Burton, H.M.

1985-01-01

TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

A model for pressure in an LMFBR duct due to discharge of gas from a failed fuel pin

International Nuclear Information System (INIS)

Srinivas, S.; Chopra, P.S.

1977-01-01

In this paper an analytical model for the calculation of pressure pulses in hexagonal ducts due to discharge of gas from a failed fuel pin is developed. The analysis yields the time history of the pressure pulse which can be used in the calculation of permanent deformation of the duct or in the assessment of the susceptibility of the duct to fracture. The real physical situation of gas discharging through a pin in a duct filled with liquid is complex to model. Here the phenomenon is modeled based on some reasonable assumptions. In this model the analysis is divided into two stages. In the first stage the gas expands as a spherical bubble, but the influence of the duct wall is taken into account. At the end of the first stage the spherical shape of the bubble is assumed to be lost and the gas is assumed to expand axially as a column. The analysis involves solving the continuity and momentum equations for the liquid along with the energy balance equation for the gas

Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

International Nuclear Information System (INIS)

Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

2007-01-01

The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

International Nuclear Information System (INIS)

Arthur, S.E.; McNeish, J.

2002-01-01

The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

Modeling of coated fuel particles irradiation behavior

International Nuclear Information System (INIS)

Liang Tongxiang; Phelip, M.

2006-01-01

In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10 -7 . As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

International Nuclear Information System (INIS)

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

Fuel from Wastewater - Harnessing a Potential Energy Source in Canada through the Co-location of Algae Biofuel Production to Sources of Effluent, Heat and CO2

Science.gov (United States)

Klise, G. T.; Roach, J. D.; Passell, H. D.; Moreland, B. D.; O'Leary, S. J.; Pienkos, P. T.; Whalen, J.

2010-12-01

Sandia National Laboratories is collaborating with the National Research Council (NRC) Canada and the National Renewable Energy Laboratory (NREL) to develop a decision-support model that will evaluate the tradeoffs associated with high-latitude algae biofuel production co-located with wastewater, CO2, and waste heat. This project helps Canada meet its goal of diversifying fuel sources with algae-based biofuels. The biofuel production will provide a wide range of benefits including wastewater treatment, CO2 reuse and reduction of demand for fossil-based fuels. The higher energy density in algae-based fuels gives them an advantage over crop-based biofuels as the “production” footprint required is much less, resulting in less water consumed and little, if any conversion of agricultural land from food to fuel production. Besides being a potential source for liquid fuel, algae have the potential to be used to generate electricity through the burning of dried biomass, or anaerobically digested to generate methane for electricity production. Co-locating algae production with waste streams may be crucial for making algae an economically valuable fuel source, and will certainly improve its overall ecological sustainability. The modeling process will address these questions, and others that are important to the use of water for energy production: What are the locations where all resources are co-located, and what volumes of algal biomass and oil can be produced there? In locations where co-location does not occur, what resources should be transported, and how far, while maintaining economic viability? This work is being funded through the U.S. Department of Energy (DOE) Biomass Program Office of Energy Efficiency and Renewable Energy, and is part of a larger collaborative effort that includes sampling, strain isolation, strain characterization and cultivation being performed by the NREL and Canada’s NRC. Results from the NREL / NRC collaboration including specific

Fuel performance evaluation through iodine activity monitoring

International Nuclear Information System (INIS)

Anantharaman, K.; Chandra, R.

1995-01-01

The objective of the failed fuel detection system is to keep a watch on fuel behaviour during operation. This paper describes the evaluation of fuel behaviour by monitoring the activities of various isotopes of iodine both during steady state and during a reactor shutdown. The limitations of this approach also has been explained. The monitoring of tramp uranium for different types of release, namely fixed contamination and continuous release from fuel, is also presented. (author)

Best Practices for Fuel System Contamination Detection and Remediation

Science.gov (United States)

2016-01-15

The University of Dayton Research Institute Best Practices for Fuel System Contamination Detection and Remediation Final Report Marlin D... Remediation Executive Summary: Fuel contamination is a broad term commonly applied to anything that causes a fuel test to fail quality assurance...Statement A: Approved for public release: distribution unlimited. 1 Best Practices for Fuel System Contamination Detection and Remediation Contents

Nuclear fuel assembly

International Nuclear Information System (INIS)

Borrman, B.; Nylund, O.

1984-01-01

A fuel assembly with a fuel channel which surrounds a plurality of fuel rods and which is divided, by means of a stiffening device of cruciform cross-section and four wings, into four sub-channels each of which comprises a bundle of fuel rods. Each fuel channel side has a plurality of stamped, inwardly-directed projections, arranged vertically one after the other, aid projections being welded to one and the same stiffening wing. Each one of the wall portions located between the projections defines, together with two adjacently positioned projections and a portion of the stiffening wing, a communiation opening between two bundles located on on one side each of the stiffening wing. (Author)

Heating subsurface formations by oxidizing fuel on a fuel carrier

Science.gov (United States)

Costello, Michael; Vinegar, Harold J.

2012-10-02

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

1995-01-01

The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described

Pattern fuel assembly loading system

International Nuclear Information System (INIS)

Ahmed, H.J.; Gerkey, K.S.; Miller, T.W.; Wylie, M.E.

1986-01-01

This patent describes an interactive system for facilitating preloading of fuel rods into magazines, which comprises: an operator work station adapted for positioning between a supply of fuel rods of predetermined types, and the magazine defining grid locations for a predetermined fuel assembly; display means associated with the work station; scanner means associated with the work station and adapted for reading predetermined information accompanying the fuel rods; a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; prompter/detector means associated with the frame for detecting insertion of a fuel rod into the magazine; and processing means responsive to the scanner means and the sensing means for prompting the operator via the display means to pre-load the fuel rods into desired grid locations in the magazine. An apparatus is described for facilitating pre-loading of fuel rods in predetermined grid locations of a fuel assembly loading magazine, comprising: a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; and means associated with the frame for detecting insertion of fuel rods into the magazine

Special storage of leaking fuel at Paks NPP

International Nuclear Information System (INIS)

Biro, Janos; Szőke, L.; Burján, T.; Lukács, R.; Hózer, H.

2015-01-01

In this paper the activities related with spent, hermetic as well as leaking fuel handling and storage, including: Spent fuel pool; Transportation criteria for the spent fuel assemblies and Interim spent fuel dry storage; Short-term storage in the spent fuel pool; Identification of the leaking assemblies by the TS-device; Present conception of Identification, handling of the leaking FAs; Modified transport procedure for the leaking FAs; Calculation of solved activity inside the leaking fuel rod; Solved activity limit values for the leaking FAs; Long-term storage in the interim spent Fuel dry storage are presented. At the end authors’ concluded that: 1) The leaking FA can be transported to the interim dry storage together with the other spent fuel assemblies in the transport container. 2) The transport-documentation of the leaking FA has to contain: isotope inventory, calculated solved activity values of the failed FA and the quantity of failed fuel rods. 3) Performing three leakage tests of the identified leaking FA before the transportation in the 5FP. it is useful to decrease the solved activity concentration inside the leaking FA and give additional information about the extent of the leakage. 4) We can calculate simply the solved activity of the leaking FA. 5) The modified transport procedure will have to be authorized. 6) The radiological effects of the leaking FA are negligible relative to the natural background radiation

Advances in HTGR fuel performance models

International Nuclear Information System (INIS)

Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

1985-01-01

Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

Fuel-pin cladding transient failure strain criterion

International Nuclear Information System (INIS)

Bard, F.E.; Duncan, D.R.; Hunter, C.W.

1983-01-01

A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

Development of Coated Particle Fuel Technology

International Nuclear Information System (INIS)

Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

2009-04-01

UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

Physics operating experience and fuel management of RAPS-1

International Nuclear Information System (INIS)

Nakra, A.N.; Purandare, H.D.; Srinivasan, K.R.; Rastogi, B.P.

1976-01-01

Rajasthan Atomic Power Station Unit-1 achieved criticality on August 11, 1972. Thereafter the reactor was brought to power, in November, 1972. Due to non-availability of the depleted fuel, the loading of which was necessary to obtain full power to begin with, the core was loaded with all natural uranium fuel and only 70% of the full power could be achieved. During the reactor operation for the last three years, the reactor has seen more than one effective full power year and about 1400 fresh fuel bundles have been loaded in the core. The reactor was subjected to about 150 power cycles resulting in more than 30% variation in operating power level and about 10 fuel bundles have failed. For satisfactory fuel management and refuelling decisions, a three dimensional simulator TRIVENI was developed. This was extensively tested during the start-up experiments and was found to be a satisfactory tool for day to day operation of the plant. In this paper, a brief account of analysis of the start-up experiments, approach to full power, power distortions and flux peaking, fuel management service and analysis of the failed fuel data has been given. (author)

Checklist for transition to new highway fuel(s).

Energy Technology Data Exchange (ETDEWEB)

Risch, C.; Santini, D.J. (Energy Systems)

2011-12-15

Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s) of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.

Telescope sipping the optimum fuel leak detection system

International Nuclear Information System (INIS)

Deleryd, R.

1998-01-01

The TELESCOPE Sipping technology is an evolutionary development from previous ABB fuel leak systems used in LWR reactors. The system utilizes the existing dynamics that cause numerous fission products to leak from a failed fuel rod when the fuel assembly is raised from a reactor core during core fuel alterations. The system can also be used by repair work in pool side inspection in order to detect leaking rods or to verify reconstituted assemblies as non leakers. (author)

Contribution to the study on the contamination of the primary coolant of pressurized water reactors by fission products. Application for evaluation of failed fuel elements

International Nuclear Information System (INIS)

Bourgeois, Pierre.

1979-01-01

After describing the radioactive substances present in the primary system, the consequences of this contamination are studied. The many publications relating to the emergence of fission products from the fuel pins are then analyzed. This study reaches the conclusion that none of the proposed methods makes it possible to interpret the activity determinations made in power reactors. This is why the opposite approach is adopted consisting in suggesting a model as from in-reactor measurements. This model will apply in the conditions for which it has been adjusted, i.e. at a steady nominal power in a 900 MWe pressurized water reactor. The originality of this model lies in the determination of a probable emergence of the can-fuel set which reflects the state of the can and the demonstration of a trapping of the iodine in the set, the consequence of which is a lower release than that of the rare gases in the primary system. After comparing the experimental results available in the literature with the results of the model, it is shown how to proceed in order to evaluate the number of failed fuel elements. Finally, on the basis of the experiments performed at the Grenoble Nuclear Study Centre, the mean and maximal predictable activites are calculated for load following working. The study of the transients is completed by a more accurate calculation of the iodine releases when the reactor is shut down which confirms the main hypotheses of the steady state model [fr

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

Fuel manufacturing and utilization

International Nuclear Information System (INIS)

2005-01-01

The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

Fuel assembly in a reactor

International Nuclear Information System (INIS)

Saito, Shozo; Kawahara, Akira.

1975-01-01

Object: To provide a fuel assembly in a reactor which can effectively prevent damage of the clad tube caused by mutual interference between pellets and the clad tube. Structure: A clad tube for a fuel element, which is located in the outer peripheral portion, among the fuel elements constituting fuel assemblies arranged in assembled and lattice fashion within a channel box, is increased in thickness by reducing the inside diameter thereof to be smaller than that of fuel elements internally located, thereby preventing damage of the clad tube resulting from rapid rise in output produced when control rods are removed. (Kamimura, M.)

Operational Experience of Nuclear Fuel in Finnish Nuclear Power Plants (with Emphasis on WWER Fuel)

International Nuclear Information System (INIS)

Teraesvirta, R.

2009-01-01

The four operating nuclear reactors in Finland, Loviisa-1 and -2 and Olkiluoto-1 and -2 have now operated approximately 30 years. The overall operational experience has been excellent. Load factors of all units have been for years among the highest in the world. The development of the fuel designs during the years has enabled remarkable improvement in the fuel performance in terms of burnup. Average discharge burnup has increased more than 30 percent in all Finnish reactor units. A systematic inspection of spent fuel assemblies, and especially all failed fuel assemblies, is a good and useful practise employed in Finland. A possibility to inspect the fuel on site using a pool side inspection facility is a relatively economic way to find out root causes of fuel failures and thereby facilitate developing remedies to prevent similar failures in the future

PWR fuel inspection and repair technology development in the Republic of Korea

International Nuclear Information System (INIS)

Park, J.Y.

1998-01-01

As of September 1997, 10 PWRs and 2 PHWRs generate 10,320MW electricity in Korea. And another 8 PWRs and 2 PHWRs will be constructed by 2006. These will need about 400 MTU of PWR fuels and 400 MTU of PHWR fuels. To improve average burnup, thermal power, fuel usability and plant safety, better poolside fuel service technologies are strongly recommended as well as the fuel design and fabrication technology improvements. During the last twenty years of nuclear power plant operation in Korea, more than 4,000 fuel assemblies has been used. At the site, continuous coolant activity measurement, pool-side visual inspection and ultrasonic tests have been performed. Some of the fuels are damaged or failed for various reasons. Some of the defected fuels were examined in hot cell to investigate the cause of failure. Even though 30 PWR fuel assemblies were repaired by foreign engineers, fuel inspection and repair technologies are not established yet. Various kind of design for the fuel make the inspection, repair and reconstitution equipment more complex. As a result, recently, a plant to obtain overall technology for poolside fuel inspection, failed fuel repair and reconstitution through R and D activities are set forth. (author)

Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

International Nuclear Information System (INIS)

Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F.

2015-01-01

Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F., E-mail: rfs@cdtn.br, E-mail: rtf@cdtn.br, E-mail: pfo@cdtn.br, E-mail: mas@cdtn.br, E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

Fuel reliability experience in Finland

International Nuclear Information System (INIS)

Kekkonen, L.

2015-01-01

Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

Energy Technology Data Exchange (ETDEWEB)

Liu, Hongbing, E-mail: liuhb07@mails.tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Du, Dong, E-mail: dudong@tsinghua.edu.cn [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); Huang, An; Chang, Baohua; Han, Zandong [Department of Mechanical Engineering, Tsinghua University, Beijing 100084 (China); Key Laboratory for Advanced Materials Processing Technology, Ministry of Education P. R. China, Beijing 100084 (China); He, Ayada [Shanghai Electric Power Generation Group Shanghai Generator Works, Shanghai 200240 (China)

2016-08-15

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Detection system for location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing

International Nuclear Information System (INIS)

Liu, Hongbing; Du, Dong; Huang, An; Chang, Baohua; Han, Zandong; He, Ayada

2016-01-01

Highlights: • A detection system for locations of pebbles transported in pipes is introduced. • The detection system is based on vibration signal processing, which is original. • The characteristics of the vibration signals of the pipe are analyzed. • The experiment shows that the detection results are accurate. • The research provides an important basis for the design of the reactor. - Abstract: Pebble-bed high temperature gas-cooled reactors have many advantages such as inherent safety, high efficiency, etc., and have been considered as a candidate for Generation IV nuclear reactors. During the operation of the reactor, there are thousands of fuel pebbles transported in the pipes outside the core by gravity and helium flow. The pattern of the pipes which consist of straight and arc sections is very complex. When a fuel pebble is transported, it will constantly collide with the pipes, especially in the arc sections. The collisions will lead to the vibration of the pipes. This paper aims to provide a detection system for the location of fuel pebbles transported in pipes in a pebble-bed reactor based on vibration signal processing. Before the reactor is running, the system acquires the vibration signals of several key sections by sensors. Then the frequency characteristics of the signals are obtained by joint time–frequency analysis. When the reactor is running, the system detects the signals and processes them based on their frequency characteristics in real time. According to the results of the processing, the system can correctly judge whether the fuel pebble has passed through the section and records the time of the passing. The experiment validates the accuracy and reliability of the detection results. In this way, the operational condition of the reactor can be monitored so that the normal running of the reactor can be ensured. Additionally, the detection data are of great significance to the evaluation and optimization of the reactor performance

Fuel-in-air FY07 summary report

Energy Technology Data Exchange (ETDEWEB)

Hanson, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wittman, Richard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wu, Wesley [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, Rick W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2008-01-22

Results of the testing program to determine fractional release rates and particle size distributions from failed commercial spent fuel related to the operations in the surface facility at Yucca Mountain are presented.

Performance evaluation of UO2-Zr fuel in power ramp tests

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.

1977-01-01

In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

Nuclear fuel banks

International Nuclear Information System (INIS)

Anon.

2010-01-01

In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)

Consequences of Fuel Failure on Criticality Safety of Used Nuclear Fuel

International Nuclear Information System (INIS)

Marshall, William J.; Wagner, John C.

2012-09-01

This report documents work performed for the Department of Energy's Office of Nuclear Energy (DOENE) Fuel Cycle Technologies Used Fuel Disposition Campaign to assess the impact of fuel reconfiguration due to fuel failure on the criticality safety of used nuclear fuel (UNF) in storage and transportation casks. This work was motivated by concerns related to the potential for fuel degradation during extended storage (ES) periods and transportation following ES, but has relevance to other potential causes of fuel reconfiguration. Commercial UNF in the United States is expected to remain in storage for longer periods than originally intended. Extended storage time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications for virtually all aspects of a UNF storage and transport system's performance. The potential impact of fuel reconfiguration on the safety of UNF in storage and transportation is dependent on the likelihood and extent of the fuel reconfiguration, which is not well understood and is currently an active area of research. The objective of this work is to assess and quantify the impact of postulated failed fuel configurations on the criticality safety of UNF in storage and transportation casks. Although this work is motivated by the potential for fuel degradation during ES periods and transportation following ES, it has relevance to fuel reconfiguration due to the effects of high burnup. Regardless of the ultimate disposition path, UNF will need to be transported at some point in the future. To investigate and quantify the impact of fuel reconfiguration on criticality safety limits, which are given in terms of the effective neutron multiplication factor, a set of failed fuel

Secondary fuel delivery system

Science.gov (United States)

Parker, David M.; Cai, Weidong; Garan, Daniel W.; Harris, Arthur J.

2010-02-23

A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

A nuclear fuel cycle system dynamic model for spent fuel storage options

International Nuclear Information System (INIS)

Brinton, Samuel; Kazimi, Mujid

2013-01-01

Highlights: • Used nuclear fuel management requires a dynamic system analysis study due to its socio-technical complexity. • Economic comparison of local, regional, and national storage options is limited due to the public financial information. • Local and regional options of used nuclear fuel management are found to be the most economic means of storage. - Abstract: The options for used nuclear fuel storage location and affected parameters such as economic liabilities are currently a focus of several high level studies. A variety of nuclear fuel cycle system analysis models are available for such a task. The application of nuclear fuel cycle system dynamics models for waste management options is important to life-cycle impact assessment. The recommendations of the Blue Ribbon Committee on America’s Nuclear Future led to increased focus on long periods of spent fuel storage [1]. This motivated further investigation of the location dependency of used nuclear fuel in the parameters of economics, environmental impact, and proliferation risk. Through a review of available literature and interactions with each of the programs available, comparisons of post-reactor fuel storage and handling options will be evaluated based on the aforementioned parameters and a consensus of preferred system metrics and boundary conditions will be provided. Specifically, three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module (WMM) which provides an easy to use interface for education on fuel cycle waste management economic impacts. Initial results of baseline cases point to positive benefits of regional storage locations with local regional storage options continuing to offer the lowest cost

Failure position detection device for nuclear fuel rod

International Nuclear Information System (INIS)

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

1987-01-01

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

Failure position detection device for nuclear fuel rod

Energy Technology Data Exchange (ETDEWEB)

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

1987-03-24

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

Axial location of cladding failure during a slow transient overpower TREAT test

International Nuclear Information System (INIS)

Page, R.J.; Murphy, W.F.; Holland, J.W.

1983-01-01

The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examination work has been completed for TREAT test J1. It was found that damage to the fuel elements during the irradiation was low enough for an accurate observation of the location of cladding failure to be made for each of the seven pins

Crud deposition modeling on BWR fuel rods

International Nuclear Information System (INIS)

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

2014-01-01

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

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

Outpatient endometrial aspiration: an alternative to methotrexate for pregnancy of unknown location.

Science.gov (United States)

Insogna, Iris G; Farland, Leslie V; Missmer, Stacey A; Ginsburg, Elizabeth S; Brady, Paula C

2017-08-01

Pregnancies of unknown location with abnormal beta-human chorionic gonadotropin trends are frequently treated as presumed ectopic pregnancies with methotrexate. Preliminary data suggest that outpatient endometrial aspiration may be an effective tool to diagnose pregnancy location, while also sparing women exposure to methotrexate. The purpose of this study was to evaluate the utility of an endometrial sampling protocol for the diagnosis of pregnancies of unknown location after in vitro fertilization. A retrospective cohort study of 14,505 autologous fresh and frozen in vitro fertilization cycles from October 2007 to September 2015 was performed; 110 patients were diagnosed with pregnancy of unknown location, defined as a positive beta-human chorionic gonadotropin without ultrasound evidence of intrauterine or ectopic pregnancy and an abnormal beta-human chorionic gonadotropin trend (location, failed intrauterine pregnancy was diagnosed in 46 patients (42%), and ectopic pregnancy was diagnosed in 64 patients (58%). Clinical variables that included fresh or frozen embryo transfer, day of embryo transfer, serum beta-human chorionic gonadotropin at the time of sampling, endometrial thickness, and presence of an adnexal mass were not significantly different between patients with failed intrauterine pregnancy or ectopic pregnancy. In patients with failed intrauterine pregnancy, 100% demonstrated adequate postsampling beta-human chorionic gonadotropin declines; villi were identified in just 46% (n=21 patients). Patients with failed intrauterine pregnancy had significantly shorter time to resolution (negative serum beta-human chorionic gonadotropin) after sampling compared with patients with ectopic pregnancy (12.6 vs 26.3 days; Plocation are spared methotrexate, with a shorter time to pregnancy resolution than those who receive methotrexate. Copyright © 2017 Elsevier Inc. All rights reserved.

BNFL Springfields Fuel Division

International Nuclear Information System (INIS)

Tarkiainen, S.; Plit, H.

1998-01-01

The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-12-01

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

Nuclear reactor seismic fuel assembly grid

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The strength of a nuclear reactor fuel assembly is enhanced by increasing the crush strength of the zircaloy spacer grids which locate and support the fuel elements in the fuel assembly. Increased resistance to deformation as a result of laterally directed forces is achieved by increasing the section modulus of the perimeter strip through bending the upper and lower edges thereof inwardly. The perimeter strip is further rigidized by forming, in the central portion thereof, dimples which extend inwardly with respect to the fuel assembly. The integrity of the spacer grid may also be enhanced by providing back-up arches for some or all of the integral fuel element locating springs and the strength of the fuel assembly may be further enhanced by providing, intermediate its ends, a steel seismic grid. 13 claims, 6 figures

The characterization and monitoring of metallic fuel breaches in EBR-2

International Nuclear Information System (INIS)

Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

1991-01-01

This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

International Nuclear Information System (INIS)

Marshall, William J.; Wagner, John C.

2012-01-01

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., 45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k eff . Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Fuel cell cooler-humidifier plate

Science.gov (United States)

Vitale, Nicholas G.; Jones, Daniel O.

2000-01-01

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

Off-gas monitor system for the detection of a failed fuel cladding tube

International Nuclear Information System (INIS)

Yuasa, Yoshiyuki; Oosaki, Masahiko; Naito, Makoto.

1984-01-01

Purpose: To rapidly and reliably detect failures in a fuel cladding tube thereby prevent accidents. Method: Off-gases discharged from an air extractor are mixed at a certain ratio with cleaning water and gamma spectra for short-life radioactive rare gases and for long-life radioactive rare gases in the off-gases are measured by gamma ray detectors. The spectra are analyzed by a Pulse-height analyzer so as to quantitatively determine the nuclides of radioactive rare gases thereby calculate the release rate on each of the nuclides. Further, a central data-processing unit calculates the composition ratio of each release models about recoil, diffusion and equilibrium in the whole radioactive rare gases based on the release rate. As the total value of the release rate is rapidly increased, recoil part will be decreased suddenly and the diffusion and equilibrium parts will be increased by so much, upon failure of a fuel can; thus the failure can rapidly be detected. (Sekiya, K.)

Functional requirements for a comprehensive transportation location referencing system

Science.gov (United States)

2000-06-01

Transportation agencies manage data that is referenced in one, two, three, and four dimensions. Location referencing system (LRS) data models vary across transportation agencies and often within organizations as well. This has resulted in failed atte...

In situ repair of a failed compression fitting

Science.gov (United States)

Wolbert, Ronald R.; Jandrasits, Walter G.

1986-01-01

A method and apparatus for the in situ repair of a failed compression fitg is provided. Initially, a portion of a guide tube is inserted coaxially in the bore of the compression fitting and locked therein. A close fit dethreading device is then coaxially mounted on the guide tube to cut the threads from the fitting. Thereafter, the dethreading device and guide tube are removed and a new fitting is inserted onto the dethreaded fitting with the body of the new fitting overlaying the dethreaded portion. Finally, the main body of the new fitting is welded to the main body of the old fitting whereby a new threaded portion of the replacement fitting is precisely coaxial with the old threaded portion. If needed, a bushing is located on the dethreaded portion which is sized to fit snugly between the dethreaded portion and the new fitting. Preferably, the dethreading device includes a cutting tool which is moved incrementally in a radial direction whereby the threads are cut from the threaded portion of the failed fitting in increments.

Engaging Future Failing States

Science.gov (United States)

2011-03-23

military missions in the Middle East, the Balkans, Africa, Asia , and South America. There is an increasing proliferation of failed and failing states...disparity, overpopulation , food security, health services availability, migration pressures, environmental degradation, personal and 22 community

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

Effects of AFR storage location on spent fuel transportation

International Nuclear Information System (INIS)

Joy, D.S.; Shappert, L.B.

1979-01-01

In order to assess the impact of Away-From-Reactor (AFR) siting on the spent fuel transportation system, five different sites were studied: Argonne, Oak Ridge, Savannah River, Idaho Falls, and Richland. Transportation costs, cask fleet sizes, and radiation exposures received by transportation workers and the general public were calculated for each site. Results show that the eastern three sites are best. 5 figures, 5 tables

Fuel exchange device for FBR type reactor

International Nuclear Information System (INIS)

Onuki, Koji.

1993-01-01

The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

International Nuclear Information System (INIS)

Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T.; Rodrigues, Juliano S.

2017-01-01

Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

Energy Technology Data Exchange (ETDEWEB)

Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T. [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Rodrigues, Juliano S., E-mail: rfs@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: rtf@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

2017-07-01

Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1989-01-01

Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

Suspension scheme for fuel pin

International Nuclear Information System (INIS)

Butts, C.E.; Gray, H.C.

1975-01-01

A description is presented of a nuclear fuel pin suspension arrangement comprising, in combination, a rod; a first beam member connected to said rod at one end; a plurality of parallel-spaced slidable fuel support plates attached to said first beam member, the longitudinal axis of first beam member being perpendicular to the longitudinal axis of each of said fuel support plates, a first coupling means disposed along the length of the first beam member for permitting slidable fuel support plates parallel movement with respect to the longitudinal axis of said first beam member, a second coupling means located at one end of each of slidable fuel plates for slidably engaging first coupling means of first beam member, a second beam member connected to the other end of each of parallel-spaced slidable fuel support plates and providing an extension, second beam member being provided with a third coupling means disposed along the length of second beam member at one end thereof; and a plurality of fuel pins provided with a fourth coupling means located at one end of each fuel pin for slidably engaging third coupling means of second beam member to permit each fuel pin parallel movement with respect to the longitudinal axis of second beam member. (U.S.)

The role of fission products (noble metal particles) in spent fuel corrosion process in a failed container

Energy Technology Data Exchange (ETDEWEB)

Wu, L., E-mail: lwu59@uwo.ca [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Univ. of Western Ontario, Surface Science Western, London, Ontario (Canada)

2013-07-01

The corrosion/dissolution of simulated spent fuel has been studied electrochemically. Fission products within the UO{sub 2} matrix are found to have significant effect on the anodic dissolution behaviour of the fuel. It is observed that H{sub 2}O{sub 2}oxidation is accelerated on the surfaces of doped noble metal (ε) particles existing in the fuel matrix. It is concluded that H{sub 2}O{sub 2} decomposition rather than UO{sub 2} corrosion should be the dominant reaction under high H{sub 2}O{sub 2} concentrations. (author)

Fuel canister and blockage pin fabrication for SLSF Experiment P4

International Nuclear Information System (INIS)

Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

1983-01-01

As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

Fuel pin response to an overpower transient in an LMFBR

International Nuclear Information System (INIS)

Grosberg, A.J.; Head, J.L.

1979-01-01

This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

Method to mount defect fuel elements i transport casks

International Nuclear Information System (INIS)

Borgers, H.; Deleryd, R.

1996-01-01

Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

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)

49 CFR 393.65 - All fuel systems.

Science.gov (United States)

2010-10-01

... motor vehicles or for the operation of auxiliary equipment installed on, or used in connection with, motor vehicles. (b) Location. Each fuel system must be located on the motor vehicle so that— (1) No part... will not contact any part of the exhaust or electrical systems of the vehicle, except the fuel level...

Navy fuel cell demonstration project.

Energy Technology Data Exchange (ETDEWEB)

Black, Billy D.; Akhil, Abbas Ali

2008-08-01

This is the final report on a field evaluation by the Department of the Navy of twenty 5-kW PEM fuel cells carried out during 2004 and 2005 at five Navy sites located in New York, California, and Hawaii. The key objective of the effort was to obtain an engineering assessment of their military applications. Particular issues of interest were fuel cell cost, performance, reliability, and the readiness of commercial fuel cells for use as a standalone (grid-independent) power option. Two corollary objectives of the demonstration were to promote technological advances and to improve fuel performance and reliability. From a cost perspective, the capital cost of PEM fuel cells at this stage of their development is high compared to other power generation technologies. Sandia National Laboratories technical recommendation to the Navy is to remain involved in evaluating successive generations of this technology, particularly in locations with greater environmental extremes, and it encourages their increased use by the Navy.

IFR fuel cycle--pyroprocess development

International Nuclear Information System (INIS)

Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

1992-01-01

The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

System of leak inspection of irradiated fuel

International Nuclear Information System (INIS)

Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F.

2007-01-01

The International Atomic Energy Agency (IAEA) through the project RLA/04/18 Irradiated Fuel Management in Research reactors, recommended among other that the participant countries (Brazil, Argentina, Chile, Peru and Mexico), develop the sipping tool to generate registrations of the state that keep the irradiated fuels in the facilities of each country. The TRIGA Mark lll Reactor (RTMIII) Department, generated a project that it is based on the dimensions of the used fuel by the RTMIII, for design and to build an inspection system of irradiated fuel well known as SIPPING. This technique, provides a high grade of accuracy in the detection of gassy fission products or liquids that escape from the enveloping of fuels that have flaws or flights. The operation process of the SIPPING is carried out generating the migration of fission products through the creation of a pressure differential gas or vacuum to identify fuel assemblies failed by means of the detection of the xenon and/or krypton presence. The SIPPING system, is a device in revolver form with 4 tangential nozzles, which will discharge the fluid between the external surface of the enveloping of the fuel and the interior surface of the encircling one; the device was designed with independent pieces, with threaded joining and with stamps to impede flights of the fluid toward the exterior of the system. The System homogenizes and it distributes the fluid pressure so that the 4 nozzles work to equality of conditions, for what the device was designed in 3 pieces, an internal that is denominated revolver, one external that calls cover, and a joining called mamelon that will unite with the main encircling of the system. The detection of fission products in failed fuels, its require that inside the encircling one where the irradiated fuel element is introduced, be generated a pressure differential of gas or vacuum, and that it allows the samples extraction of water. For what generated a top for the encircling with the

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

Stress and Diffusion in Stored Pu ZPPR Fuel from Alpha Generation

Energy Technology Data Exchange (ETDEWEB)

Charles W. Solbrig; Chad L. Pope; Jason P. Andrus

2014-07-01

ZPPR (Zero Power Physics Reactor) is a research reactor that has been used to investigate breeder reactor fuel designs. The reactor has been dismantled but its fuel is still stored there. Of concern are its plutonium containing metal fuel elements which are enclosed in stainless steel cladding with gas space filled with helium–argon gas and welded air tight. The fuel elements which are 5.08 cm by 0.508 cm up to 20.32 cm long (2 in × 0.2 in × 8 in) were manufactured in 1968. A few of these fuel elements have failed releasing contamination raising concern about the general state of the large number of other fuel elements. Inspection of the large number of fuel elements could lead to contamination release so analytical studies have been conducted to estimate the probability of failed fuel elements. This paper investigates the possible fuel failures due to generation of helium in the metal fuel from the decay of Pu and its possible damage to the fuel cladding from metal fuel expansion or from diffusion of helium into the fuel gas space. This paper (1) calculates the initial gas loading in a fuel element and its internal free volume after it has been brought into the atmosphere at ZPPR, (2) shows that the amount of helium generated by decay of Pu over 46 years since manufacture is significantly greater than this initial loading, (3) determines the amount of fuel swelling if the helium stays fixed in the fuel plate and estimates the amount of helium which diffuses out of the fuel plate into the fuel plenum assuming the helium does not remain fixed in the fuel plate but can diffuse to the plenum and possibly through the cladding. Since the literature is not clear as to which possibility occurs, as with Schroedinger’s cat, both possibilities are analyzed. The paper concludes that (1) if the gas generated is fixed in the fuel, then the fuel swelling it can cause would not cause any fuel failure and (2) if the helium does diffuse out of the fuel (in accordance

Paper summary inventory assessment of DOE spent nuclear fuels

International Nuclear Information System (INIS)

Abbott, D.G.; Bringhurst, A.R.; Fillmore, D.L.

1994-01-01

The U.S. Department of Energy (DOE) has determined that it will not longer reprocess its spent nuclear fuel. This decision made it necessary to manage this fuel for long-term interim storage and ultimate disposal. DOE is developing a computerized database of its spent nuclear fuel inventory. This database contains information about the fuels and the fuel storage locations. There is approximately 2,618 metric tons initial heavy metal of fuel, stored at 12 locations. For analysis in an environmental impact statement, the fuel has been divided into six categories: naval, aluminum-based, Hanford defense, graphite, commercial-type, and test and experimental. This paper provides a discussion of the development of the database, and includes summary inventory information and a brief description of the fuels

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

International Nuclear Information System (INIS)

Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-01-01

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-04-15

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Dams designed to fail

Energy Technology Data Exchange (ETDEWEB)

Penman, A. [Geotechnical Engineering Consultants, Harpenden (United Kingdom)

2004-09-01

New developments in geotechnical engineering have led to methods for designing and constructing safe embankment dams. Failed dams can be categorized as those designed to fail, and those that have failed unexpectedly. This presentation outlined 3 dam failures: the 61 m high Malpasset Dam in France in 1959 which killed 421; the 71 m high Baldwin Hills Dam in the United States in 1963 which killed 5; and, the Vajont Dam in Italy in 1963 which killed 2,600 people. Following these incidents, the International Commission for Large Dams (ICOLD) reviewed regulations on reservoir safety. The 3 dams were found to have inadequate spillways and their failures were due to faults in their design. Fuse plug spillways, which address this problem, are designed to fail if an existing spillway proves inadequate. They allow additional discharge to prevent overtopping of the embankment dam. This solution can only be used if there is an adjacent valley to take the additional discharge. Examples of fuse gates were presented along with their effect on dam safety. A research program is currently underway in Norway in which high embankment dams are being studied for overtopping failure and failure due to internal erosion. Internal erosion has been the main reason why dams have failed unexpectedly. To prevent failures, designers suggested the use of a clay blanket placed under the upstream shoulder. However, for dams with soft clay cores, these underblankets could provide a route for a slip surface and that could lead to failure of the upstream shoulder. It was concluded that a safe arrangement for embankment dams includes the use of tipping gates or overturning gates which always fail at a required flood water level. Many have been installed in old and new dams around the world. 14 refs., 19 figs.

Fatty Acid Oxidation Is Preserved Regardless of Impaired Uptake in the Chronically Failing Rat Heart

OpenAIRE

TACHIKAWA, Hitoshi

2004-01-01

Fatty acid is used as a major fuel in the fasting heart, but the precise metabolism in the failing heart remains unknown. We assessed the hypothesis that the fatty acid metabolism might be impaired or delayed during heart failure. We examined in vivo kinetics of an isotope-labeled fatty acid analogue and its substrates as well as hemodynamic parameters and histopathological findings in a rat model of postmyocarditic dilated cardiomyopathy. Rat experimental autoimmune myocarditis (EAM) was ind...

Nuclear fuel assembly

International Nuclear Information System (INIS)

Betten, P.R.

1976-01-01

Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Failed endotracheal intubation

Directory of Open Access Journals (Sweden)

Sheykhol Islami V

1995-07-01

Full Text Available The incidence of failed intubation is higher in obstetric than other surgical patients. Failed intubation was the 2nd commonest cause of mortality during anesthesia. Bearing in mind that failre to intubate may be unavoidable in certain circumstances, it is worth reviewing. The factors, which may contribute to a disastrous out come. Priorities of subsequent management must include maintaining oxygenation and preventing aspiration of gastric contents. Fiber optic intubation is now the technique of choice with a high success rate and with least trauma to the patient.

Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

International Nuclear Information System (INIS)

Misfeldt, I.

1980-01-01

A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

Development of evaluation method of fuel failure fraction during the High Temperature Engineering Test Reactor operation

Energy Technology Data Exchange (ETDEWEB)

Sawa, Kazuhiro; Yoshimuta, Shigeharu; Tobita, Tsutomu; Sato, Masashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1997-05-01

The High Temperature Engineering Test Reactor (HTTR) uses coated particles as fuel. During normal operation, short-lived noble gases are mainly released by diffusion from fuel particles with defects in their coating layers (i.e., failed particle). Since noble gases do not plate out on the inner surfaces of primary cooling system, their activities in primary coolant reflect fuel failure fraction in the core. An evaluation method was developed to predict failure fraction of coated fuel particles during normal operation of the HTTR. The method predicts core-average and hot plenum regionwise failure fractions based on the fractional releases, (R/B)s, of noble gases. The (R/B)s are calculated by fission gas concentration measurements in the primary cooling system of the HTTR. Recent fabrication data show that through-coatings failure fraction is extremely low. Then, fractional release from matrix contamination uranium, which is background for accurate evaluation of the fuel failure fraction, should be precisely predicted. This report describes an evaluation method of fuel failure fraction from measurements in the HTTR together with a fission gas release model from fuel compact containing failed particles and matrix contamination uranium. (author)

Radial power distribution shaping within a PWR fuel assembly utilizing asymmetrically loaded gadolinia-bearing fuel pins

International Nuclear Information System (INIS)

Stone, I.Z.

1992-01-01

As in-core fuel management designs evolve to meet the demands of increasing energy output, more innovative methods are developed to maintain power peaking within acceptable thermal margin limits. In-core fuel management staff must utilize various loading pattern strategies such as cross-core movement of fuel assemblies, multibatch enrichment schemes, and burnable absorbers as the primary means of controlling the radial power distribution. The utilization of fresh asymmetrically loaded gadolinia-bearing assemblies as a fuel management tool provides an additional means of controlling the radial power distribution. At Siemens Nuclear Power Corporation (SNP), fresh fuel assemblies fabricated with asymmetrically loaded gadolinia-bearing fuel rods have been used successfully for several cycles of reactor operation. Asymmetric assemblies are neutronically modeled using the same tools and models that SNP uses to model symmetrically loaded gadolinia-bearing fuel assemblies. The CASMO-2E code is used to produce the homogenized macroscopic assembly cross sections for the nodal core simulator. Optimum fuel pin locations within the asymmetrical assembly are determined using the pin-by-pin PDQ7 assembly core model for each new assembly design. The optimum pin location is determined by the rod loading that minimizes the peak-to-average pin power

Ducted fuel injection

Energy Technology Data Exchange (ETDEWEB)

Mueller, Charles J.

2018-03-06

Various technologies presented herein relate to enhancing mixing inside a combustion chamber to form one or more locally premixed mixtures comprising fuel and charge-gas with low peak fuel to charge-gas ratios to enable minimal, or no, generation of soot and other undesired emissions during ignition and subsequent combustion of the locally premixed mixtures. To enable sufficient mixing of the fuel and charge-gas, a jet of fuel can be directed to pass through a bore of a duct causing charge-gas to be drawn into the bore creating turbulence to mix the fuel and the drawn charge-gas. The duct can be located proximate to an opening in a tip of a fuel injector. The duct can comprise of one or more holes along its length to enable charge-gas to be drawn into the bore, and further, the duct can cool the fuel and/or charge-gas prior to combustion.

Evaluation of the effect of probe design parameters on ECT signal and development of eddy current probe for irradiated fuel rods

International Nuclear Information System (INIS)

Kwank, S. W.; Han, Y. K.; Woo, S. K.; Kim, T. W.; Park, J. Y.; Kim, B. J.; Park, J. Y.

1999-01-01

Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dolbow, John E. [Duke Univ., Durham, NC (United States); Hales, Jason D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes

Improved Direct Methanol Fuel Cell Stack

Science.gov (United States)

Wilson, Mahlon S.; Ramsey, John C.

2005-03-08

A stack of direct methanol fuel cells exhibiting a circular footprint. A cathode and anode manifold, tie-bolt penetrations and tie-bolts are located within the circular footprint. Each fuel cell uses two graphite-based plates. One plate includes a cathode active area that is defined by serpentine channels connecting the inlet and outlet cathode manifold. The other plate includes an anode active area defined by serpentine channels connecting the inlet and outlet of the anode manifold, where the serpentine channels of the anode are orthogonal to the serpentine channels of the cathode. Located between the two plates is the fuel cell active region.

Fuel Retrieval and Management of Fuel Element Debris

International Nuclear Information System (INIS)

Chande, Shridhar; Lachaume, J. L.

2013-01-01

Nuclear accidents involving core meltdown have not been so rare. While the first occurred in early fifties, it is reported that about 20 have occurred worldwide in military and commercial reactors. The more recent and major accidents are 1. Three Mile Island, USA in 1979: Approximately half the core was melted, and flowed to the bottom of the reactor pressure vessel however the pressure vessel remained intact and contained the damaged fuel. 2. Chernobyl, former USSR in 1984: Explosive release of radioactive material occurred. About 6 tons of fuel was dispersed as air-borne particles. Most of the core was damaged or melted. 3. Fukushima, Japan 2011: Three units suffered melt down. In unit 1 almost all the fuel assemblies melted and accumulated at the bottom of the vessel. It is reported that the vessel failed and the molten corium has penetrated the concrete. In the units 2 and 3, partial melting of cores has occurred. In several of these cases, fuel retrieval and management activities have been carried out. The experience and insights gained from these activities will be extremely useful for planning and execution of similar activities in future if ever they are needed. The purpose of this session was to exchange this experience and also to share the lessons learned. This is of particularly important, at this juncture, when planning and preparation for retrieval of damaged cores in Fukushima NPP is in progress. (author)

Handling apparatus for fuel assemblies in a core

International Nuclear Information System (INIS)

Hatakenaka, Hideo.

1975-01-01

Object: To prevent an occurrence of a cloud as well as trouble in outflow of cooling water at the time of failure, in a window through which the operation of a collet installing and removing mechanism is monitored. Structure: A monitoring window comprises a pair of transparent window panes between which is interposed a non-compressive transparent fluid. With this construction, when the collet installing and removing mechanism within a container is operated while illuminating it by light means and monitoring it by a television camera to connect a fuel assembly with a shielding plug, and even if one transparent window pane should be failed as a result of trouble, the other transparent window pane prevents outflow of cooling water within a fuel transferring transfer port, and at the same time, the scattering force of fragments of failed transparent window pane is attenuated by the non-compressive transparent body within the monitoring window chamber. (Hanada, M.)

Bio-fuels

International Nuclear Information System (INIS)

2008-01-01

This report presents an overview of the technologies which are currently used or presently developed for the production of bio-fuels in Europe and more particularly in France. After a brief history of this production since the beginning of the 20. century, the authors describe the support to agriculture and the influence of the Common Agricultural Policy, outline the influence of the present context of struggle against the greenhouse effect, and present the European legislative context. Data on the bio-fuels consumption in the European Union in 2006 are discussed. An overview of the evolution of the activity related to bio-fuels in France, indicating the locations of ethanol and bio-diesel production facilities, and the evolution of bio-fuel consumption, is given. The German situation is briefly presented. Production of ethanol by fermentation, the manufacturing of ETBE, the bio-diesel production from vegetable oils are discussed. Second generation bio-fuels are then presented (cellulose enzymatic processing), together with studies on thermochemical processes and available biomass resources

Commercialization of nuclear fuel cycle business

International Nuclear Information System (INIS)

Yakabe, Hideo

1998-01-01

Japan depends on foreign countries almost for establishing nuclear fuel cycle. Accordingly, uranium enrichment, spent fuel reprocessing and the safe treatment and disposal of radioactive waste in Japan is important for securing energy. By these means, the stable supply of enriched uranium, the rise of utilization efficiency of uranium and making nuclear power into home-produced energy can be realized. Also this contributes to the protection of earth resources and the preservation of environment. Japan Nuclear Fuel Co., Ltd. operates four business commercially in Rokkasho, Aomori Prefecture, aiming at the completion of nuclear fuel cycle by the technologies developed by Power Reactor and Nuclear Fuel Development Corporation and the introduction of technologies from foreign countries. The conditions of location of nuclear fuel cycle facilities and the course of the location in Rokkasho are described. In the site of about 740 hectares area, uranium enrichment, burying of low level radioactive waste, fuel reprocessing and high level waste control have been carried out, and three businesses except reprocessing already began the operation. The state of operation of these businesses is reported. Hereafter, efforts will be exerted to the securing of safety through trouble-free operation and cost reduction. (K.I.)

Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

International Nuclear Information System (INIS)

Peacock, H.B.

1983-08-01

Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

Location-based restoration mechanism for multi-domain GMPLS networks

DEFF Research Database (Denmark)

Manolova, Anna Vasileva; Calle, Eusibi; Ruepp, Sarah Renée

2009-01-01

In this paper we propose and evaluate the efficiency of a location-based restoration mechanism in a dynamic multi-domain GMPLS network. We focus on inter-domain link failures and utilize the correlation between the actual position of a failed link along the path with the applied restoration...

Operational reliability testing of FBR fuel in EBR-II

International Nuclear Information System (INIS)

Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

1991-01-01

The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

Experiments to investigate the effects of small changes in fuel stoichiometry on fission gas release

Energy Technology Data Exchange (ETDEWEB)

Copeland, P S; Smith, R C [Windscale Lab., AEA Technology, Seascale, Cumbria (United Kingdom)

1997-08-01

Fuel pin failure in-reactor leads to fission product and in the case of a PWR fuel debris release to the coolant. For economic reasons immediate shutdown and discharge of failed fuel needs to be avoided but this needs to be counter-balanced against the increasing dose to operators. PWR practice is to continue running wit failed rods, monitoring coolant activity, and only shutting down the reactor and discharging the fuel when circuit activity levels become unacceptable. The rate of fission product release under failed fuel conditions is of key importance and considerable effort has been directed towards establishing the dependency of release on temperature, heating rate, burn-up, and also the extent of fuel oxidation. As a precursor to a possible wider investigation of this area, a small programme was mounted during 1992/1993 to confirm whether small changes in the oxidation state of the fuel, for example those caused by minor cladding defects, would significantly effect fuel behaviour during postulated design basis faults. The objective of the programme was to determine the effects of small departures from stoichiometric fuel composition on fission gas release, and to compare the results with the current methodology for calculating releases under fault conditions. A total of eight experiments was performed. Two were intended as baseline tests to provide a reference with which to compare the effect of oxidation state influenced behaviour with that of thermal effects. It was found that small changes in stoichiometry of {sup {approx}}1 x 10{sup -6} had little or no effect on release but that changes of {sup {approx}} 1 x 10{sup -4} were observed to increase the diffusion coefficient, for {sup 85}Kr, by up to an order of magnitude and hence greatly increase the release rate. The stoichiometry of the sample used in these tests was, for convenience, adjusted using He/H{sub 2}/H{sub 2}O atmospheres. (Abstract Truncated)

Determining fissile content of nuclear fuel elements

International Nuclear Information System (INIS)

Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

1980-01-01

This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

European experience in the transport of irradiated light-water reactor fuel

International Nuclear Information System (INIS)

Curtis, H.W.

1979-01-01

Various methods of transport of irradiated fuel flasks in Europe are described. While many problems in the transport of heavy flasks have been solved some remain and new ones have appeared. Some of these problems are the accumulation of crud on the surface of fuel elements, the problems of failed fuel, stringent criticality criteria, the ''sweating out'' of contaminated flasks, the access, road or rail, to reactor sites, and the maintenance of the transport vehicles. Some future trends in the direction of heavy flasks in the range of 75 to 100 tonnes are indicated

Performance assessment analyses unique to Department of Energy spent nuclear fuel

International Nuclear Information System (INIS)

Loo, H.H.; Duguid, J.J.

2000-01-01

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

Investigation of the ramp testing behaviour of fuel pins with different diameters

International Nuclear Information System (INIS)

Pott, G.; Herren, M.; Wigger, B.

1979-09-01

The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

Fact sheet on fuel manufacturing and utilization

International Nuclear Information System (INIS)

2006-01-01

The Nuclear Fuel Cycle and Materials Section (NFCMS) supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle, provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection, supports the development fuel modeling expertise in Member States, covering both normal operation and postulated and severe accident conditions, provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation, supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, provides information and support research into the basic properties of fuel materials, including UO2, MOX, (Th, Pu)O2, (Th, U233)O2 fuels and zirconium alloy cladding and fuel assembly components and offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology and materials, economic and other aspects of nuclear fuel use (e.g. environmental impact). Recently NFCMS provided support to a Member State manufacturing Gadolinia doped fuel and provided in-mast sipping equipment to a Nuclear Power Plant to allow the determination of fuel failure. Member States interested in fuel performance and manufacture should contact the Technical Cooperation Department of the Agency and Member States interested in knowing more about the Agency's programme on source management should contact: C. Ganguly, Section Head, V. Inozemtsev, J. Killeen

Drying apparatus located back and forth of a Voloxidizer for reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Harada, Keizo.

1975-01-01

Object: To permit a continuous spent fuel drying operation during a thermal oxidation process (voroxidation) effected between a shearing step and dissolution step in fuel retreatment. Structure: The invention comprises a gas-tight cylindrical outer barrel, a rotative shaft hermetically supported in the outer barrel, a cylindrical inner barrel having at least three chambers secured to the rotative shaft, a fuel inlet nozzle provided at a position corresponding to the top of the outer barrel and one of the chambers of the inner barrel, a fuel outlet nozzle provided at the bottom of the outer barrel and at the bottom of one of the other chambers of the inner barrel, and a drying gas inlet nozzle and an off-gas outlet nozzle provided at positions corresponding to the side wall of the outer barrel and the side wall of one the other chambers of the inner barrel. The spent fuel is continuously dried by intermittently rotating the inner barrel. (Kamimura, M.)

33 CFR 183.564 - Fuel tank fill system.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel tank fill system. 183.564...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.564 Fuel tank fill system. (a) Each fuel fill opening must be located so that a gasoline overflow of up to five...

Characterization of the 309 building fuel transfer pit and storage basin

International Nuclear Information System (INIS)

Hale, N.S.

1998-01-01

This document identifies radiological, chemical and physical conditions inside the Fuel Transfer Pit and Fuel Storage Basins. These spaces are located inside the Plutonium Recycle Test Reactor structure (309 Building.) The fuel handling and storage feature of the PRTR were primarily located in these spaces. The conditions were assessed as part of overall 309 Building transition

Critical thinking: are the ideals of OBE failing us or are we failing the ...

African Journals Online (AJOL)

Critical thinking: are the ideals of OBE failing us or are we failing the ideals of OBE? K Lombard, M Grosser. Abstract. No Abstract. South African Journal of Education Vol. 28 (4) 2008: pp. 561-580. Full Text: EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT.

Wet motor geroter fuel pump

Energy Technology Data Exchange (ETDEWEB)

Wiernicki, M.V.

1987-05-05

This patent describes a wet motor gerotor fuel pump for pumping fuel from a fuel source to an internal combustion which consists of: gerotor pump means comprising an inner pump gear, an outer pump gear, and second tang means located on one of the inner and outer pump gears. The second tang means further extends in a second radial direction radially offset from the first radial direction and forms a driving connection with the first tang means such that the fuel pump pumps fuel from the fuel source into the narrow conduit inlet chamber, through the gerotor pump means past the electric motor means into the outlet housing means substantially along the flow axis to the internal combustion engine.

Fusion fuel cycle solid radioactive wastes

International Nuclear Information System (INIS)

Gore, B.F.; Kaser, J.D.; Kabele, T.J.

1978-06-01

Eight conceptual deuterium-tritium fueled fusion power plant designs have been analyzed to identify waste sources, materials and quantities. All plant designs include the entire D-T fuel cycle within each plant. Wastes identified include radiation-damaged structural, moderating, and fertile materials; getter materials for removing corrosion products and other impurities from coolants; absorbents for removing tritium from ventilation air; getter materials for tritium recovery from fertile materials; vacuum pump oil and mercury sludge; failed equipment; decontamination wastes; and laundry waste. Radioactivity in these materials results primarily from neutron activation and from tritium contamination. For the designs analyzed annual radwaste volume was estimated to be 150 to 600 m 3 /GWe. This may be compared to 500 to 1300 m 3 /GWe estimated for the LMFBR fuel cycle. Major waste sources are replaced reactor structures and decontamination waste

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

Thermal breeder fuel enrichment zoning

International Nuclear Information System (INIS)

Capossela, H.J.; Dwyer, J.R.; Luce, R.G.; McCoy, D.F.; Merriman, F.C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect. 1 figure

Numerical analyses of an ex-core fuel incident: Results of the OECD-IAEA Paks Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hozer, Z., E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Aszodi, A. [BME NTI Budapest (Hungary); Barnak, M. [IVS, Trnava (Slovakia); Boros, I. [BME NTI Budapest (Hungary); Fogel, M. [VUJE, Trnava (Slovakia); Guillard, V. [IRSN, Cadarache (France); Gyori, Cs. [ITU, EU, Karlsruhe (Germany); Hegyi, G. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Horvath, G.L. [VEIKI, Budapest (Hungary); Nagy, I. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Junninen, P. [VTT, Espoo (Finland); Kobzar, V. [KI, Moscow (Russian Federation); Legradi, G. [BME NTI Budapest (Hungary); Molnar, A. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Pietarinen, K. [VTT, Espoo (Finland); Perneczky, L. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Makihara, Y. [ATMEA, Paris (France); Matejovic, P. [IVS, Trnava (Slovakia); Perez-Fero, E.; Slonszki, E. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary)

2010-03-15

The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data. The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 deg. C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg.

46 CFR 58.50-5 - Gasoline fuel tanks.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Gasoline fuel tanks. 58.50-5 Section 58.50-5 Shipping... AND RELATED SYSTEMS Independent Fuel Tanks § 58.50-5 Gasoline fuel tanks. (a) Construction—(1) Shape...) Installation. (1) Gasoline fuel tanks used for propulsion shall be located in water-tight compartments separate...

Evaluation of fuel release rate and mechanism tests under RBCB conditions

International Nuclear Information System (INIS)

Adamson, M.G.

1981-09-01

This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory

Tennessee Valley Authority nuclear fuel performance

International Nuclear Information System (INIS)

Rahn, B.; Gardner, D.

1985-01-01

The Browns Ferry BWR units started operation in the early to mid 1970s, and the Sequoyah PWR units started in the very early 1980s. Operating dates, outage length, capacity factor, and cycle exposure are given for these units. This paper deals with fuel performance at the Browns Ferry and Sequoyah plants. Leaks, off-gas, fraction failed, iodine-131, and corrosion are covered. 14 figs

Ex-vessel nuclear fuel transfer system

International Nuclear Information System (INIS)

Wade, E.E.

1978-01-01

A system for transferring fuel assemblies between a fuel transfer area and a fuel storage area while the fuel assemblies remain completely submerged in a continuous body of coolant is described. A fuel transfer area filled with reactor coolant communicating with the reactor vessel below the reactor coolant level provides a transfer area for fuel assemblies in transit to and from the reactor vessel. A positioning mechanism comprising at least one rotatable plug disposed on a fuel transfer tank located outside the reactor vessel cooperates with either the fuel transfer area or the fuel storage area to position a fuel assembly in transit. When in position, a transporting mechanism cooperating with the positioning mechanism lifts or lowers a chosen fuel assembly. The transporting mechanism together with the positioning mechanism are capable of transferring a fuel assembly between the fuel transfer area and the fuel storage area

Elastic-plastic stress distributions near the endcap of a fuel element

International Nuclear Information System (INIS)

Tayal, M.; Hallgrimson, K.D.; Sejnoha, R.; Singh, P.N.

1993-06-01

This paper discusses the stress patterns in and near the endcap of a CANDU fuel element from the perspective of stress corrosion cracking. Simulations of out-reactor burst tests suggest that local plastic strains stay comparatively low for internal pressures below 26-30 MPa. Photoelastic measurements as well as analytical assessments show that the reentrant corner at the sheath/endcap junction results in high concentration of stresses and strains. Analytical assessments show that the in-reactor stresses and strains at the reentrant corner are highly multiaxial, and well into the plastic range. The maximum principal stress correlates well with the location and the direction of circumferential endcap cracks observed in fuel that failed in the Bruce reactor. Thus the maximum principal stress appears promising in ranking various geometries of the sheath/endcap junction with respect to their relative susceptibility to stress corrosion cracking. Design studies suggest that the most effective practical ways of lowering the stresses near the weld, in order of decreasing importance, are to provide a larger interference-free length between the ridge and the endcaps; to increase the pellet/sheath radial gap; to increase the pellet/endcap axial gap; and to keep the gas pressure low. (author). 16 refs., 16 figs

NAC international dry spent fuel transfer technology

International Nuclear Information System (INIS)

Shelton, Thomas A.; Malone, James P.; Patterson, John R.

1996-01-01

Full text: For more than ten years NAC International (NAC) has designed, fabricated, tested and operated a variety of Dry Transfer Systems (DTS's) to transfer spent nuclear fuel from facilities with limited crane capabilities, limited accesses or limiting features to IAEA and USNRC licensed spent fuel transport casks or vice-versa. These DTS's have been operated in diverse environments in the United States and throughout the world and have proven to be a significant enhancement in transferring fuel between spent fuel pools, dry storage and hot cell facilities and spent fuel transport casks. Over the years, NAC has successfully and safely transferred more than two thousand fuel assemblies in DTS's. Our latest generation DTS incorporates years of extensive design and operating experience. It consists of a transfer cask with integrated fuel canister grapple, fuel canisters, and facility and cask adapters as well as a complement of related tools and equipment. The transfer cask is used to move irradiated HEU and LEU MTR fuel onsite in those instances where direct loading or unloading of the shipping cask is not possible due to dimensional, weight or other restrictions. The transfer cask is used to move canisters of fuel from the fuel storage location to the shipping cask. Adapters are employed to ensure proper interfacing of the transfer cask with fuel storage locations and shipping casks (NAC-LWT and NLI-1/2). Our existing fuel storage location adapter is designed for use with a storage pool; however, site or equipment specific adapters can easily be developed to allow interfacing with virtually any storage facility. Prior to movement of the first fuel canister in the transfer cask, the shipping cask is prepared for loading by proper set up of the base plate, shipping cask and shipping cask adapter. The fuel canisters are loaded with fuel and then retracted into the transfer cask via the fuel storage location adapter. The transfer cask is then moved to the shipping

Assesment of strength and integrity of fuel channels

International Nuclear Information System (INIS)

2000-01-01

Detailed analysis to base strength and integrity of fuel channels was necessary for the licensing process. Description of tasks performed in this direction in 1999 is presented: fuel channel independent strength calculations, assessment of present fuel channels state, analysis of dynamic processes during partial group distribution header rupture, structural integrity analysis of fuel channels located next to broke channel

Light-duty vehicle fuel economy improvements, 1979--1998: A consumer purchase model of corporate average fuel economy, fuel price, and income effects

Science.gov (United States)

Chien, David Michael

2000-10-01

The Energy Policy and Conservation Act of 1975, which created fuel economy standards for automobiles and light trucks, was passed by Congress in response to the rapid rise in world oil prices as a result of the 1973 oil crisis. The standards were first implemented in 1978 for automobiles and 1979 for light trucks, and began with initial standards of 18 MPG for automobiles and 17.2 MPG for light trucks. The current fuel economy standards for 1998 have been held constant at 27.5 MPG for automobiles and 20.5 MPG for light trucks since 1990--1991. While actual new automobile fuel economy has almost doubled from 14 MPG in 1974 to 27.2 MPG in 1994, it is reasonable to ask if the CAFE standards are still needed. Each year Congress attempts to pass another increase in the Corporate Average Fuel Economy (CAFE) standard and fails. Many have called for the abolition of CAFE standards citing the ineffectiveness of the standards in the past. In order to determine whether CAFE standards should be increased, held constant, or repealed, an evaluation of the effectiveness of the CAFE standards to date must be established. Because fuel prices were rising concurrently with the CAFE standards, many authors have attributed the rapid rise in new car fuel economy solely to fuel prices. The purpose of this dissertation is to re-examine the determinants of new car fuel economy via three effects: CAFE regulations, fuel price, and income effects. By measuring the marginal effects of the three fuel economy determinants upon consumers and manufacturers choices, for fuel economy, an estimate was made of the influence of each upon new fuel economy. The conclusions of this dissertation present some clear signals to policymakers: CAFE standards have been very effective in increasing fuel economy from 1979 to 1998. Furthermore, they have been the main cause of fuel economy improvement, with income being a much smaller component. Furthermore, this dissertation has suggested that fuel prices have

Characterization of the 309 fuel examination facility

International Nuclear Information System (INIS)

Greenhalgh, W.O.; Cornwell, B.C.

1997-01-01

This document identifies radiological, chemical and physical conditions inside the Fuel Examination Facility. It is located inside the Plutonium Recycle Test Reactor containment structure (309 Building.) The facility was a hot cell used for examination of PRTR fuel and equipment during the 1960's. Located inside the cell is a PRTR shim rod assembly, reported are radiological conditions of the sample. The conditions were assessed as part of overall 309 Building transition

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

1986-01-01

In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

Identical location transmission electron microscopy in combination with rotating disc electrode measurements. The activity of fuel cell catalysts and their degradation

Energy Technology Data Exchange (ETDEWEB)

Schloegl, Katrin G.

2011-07-13

As an alternative to conventional combustion engines, the Proton Exchange Membrane Fuel Cell (PEMFC) using hydrogen as a fuel is a promising concept owing to its potential independence from fossil fuels, high efficiency and zero emissions. Concerning its commercial viability, the fundamental problem of high system cost per power output and lifetime is closely related to finding more active and stable catalysts for the oxygen reduction reaction. In the presented work, several methods are combined to examine the parameters and processes responsible for both activity and degradation of platinum-based catalysts. Degradation mechanisms are scrutinized by means of electrochemical measurements with the rotating disc electrode in combination with a recently developed TEM technique, which allows for the comparison of identical locations before and after accelerated stress tests. (orig.) [German] Die mit Wasserstoff betriebene Proton Exchange Membrane Brennstoffzelle (PEMFC) stellt aufgrund ihrer potentiellen Unabhaengigkeit von fossilen Energietraegern, ihrem hohen Wirkungsgrad und fehlendem Schadstoffausstoss eine vielversprechende Alternative zum konventionellen Verbrennungsmotor dar. Das grundlegende Problem der zu hohen Systemkosten und zu geringen Lebensdauer fuer kommerzielle Anwendungen ist eng mit der Entwicklung aktiverer und stabiler Elektrokatalysatoren fuer die Sauerstoffreduktion verknuepft. In der vorliegenden Arbeit werden verschiedene Methoden kombiniert, um die Parameter und Prozesse zu untersuchen, welche fuer die Aktivitaet und Degradation platinbasierter Katalysatoren verantwortlich sind. Zur Aufklaerung vorliegender Degradationsmechanismen werden elektrochemische Messungen mit der rotierenden Scheibenelektrode in Kombination mit einer neu entwickelten TEM Methode eingesetzt, welche es ermoeglicht, identische Stellen vor und nach beschleunigten Degradationstests zu untersuchen.

Postirradiation examination of Peach Bottom HTGR Driver Fuel Element E06-01

International Nuclear Information System (INIS)

Dyer, F.F.; Wichner, R.P.; Martin, W.J.; Fairchild, L.L.; Kedl, R.J.; de Nordwall, H.J.

1976-04-01

The report presented describes the postirradiation examinations of driver fuel element E06-01, which had been irradiated an equivalent of 384 full-power days in Peach Bottom, Unit 1. The fuel element is described in detail and its temperature and irradiation service history briefly outlined. Results presented include: (1) visual observations; (2) critical dimensions of fuel compacts, sleeve, and spine; (3) axial distributions of gamma-emitting nuclides plus 3 H and 90 Sr; (4) radial distributions of these nuclides in the sleeve and spine at three axial locations in the fueled regions and three locations in the upper reflector; (5) metallographic examination of samples of fuel compact material; and (6) burnup determinations via radiochemical analyses at two compact locations

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

Failed rib region prediction in a human body model during crash events with precrash braking.

Science.gov (United States)

Guleyupoglu, B; Koya, B; Barnard, R; Gayzik, F S

2018-02-28

The objective of this study is 2-fold. We used a validated human body finite element model to study the predicted chest injury (focusing on rib fracture as a function of element strain) based on varying levels of simulated precrash braking. Furthermore, we compare deterministic and probabilistic methods of rib injury prediction in the computational model. The Global Human Body Models Consortium (GHBMC) M50-O model was gravity settled in the driver position of a generic interior equipped with an advanced 3-point belt and airbag. Twelve cases were investigated with permutations for failure, precrash braking system, and crash severity. The severities used were median (17 kph), severe (34 kph), and New Car Assessment Program (NCAP; 56.4 kph). Cases with failure enabled removed rib cortical bone elements once 1.8% effective plastic strain was exceeded. Alternatively, a probabilistic framework found in the literature was used to predict rib failure. Both the probabilistic and deterministic methods take into consideration location (anterior, lateral, and posterior). The deterministic method is based on a rubric that defines failed rib regions dependent on a threshold for contiguous failed elements. The probabilistic method depends on age-based strain and failure functions. Kinematics between both methods were similar (peak max deviation: ΔX head = 17 mm; ΔZ head = 4 mm; ΔX thorax = 5 mm; ΔZ thorax = 1 mm). Seat belt forces at the time of probabilistic failed region initiation were lower than those at deterministic failed region initiation. The probabilistic method for rib fracture predicted more failed regions in the rib (an analog for fracture) than the deterministic method in all but 1 case where they were equal. The failed region patterns between models are similar; however, there are differences that arise due to stress reduced from element elimination that cause probabilistic failed regions to continue to rise after no deterministic failed region would be

Nuclear fuels

International Nuclear Information System (INIS)

2008-01-01

The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

1994-10-01

This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

X-ray cinematography on the nuclear fuel and cladding motion diagnostics

International Nuclear Information System (INIS)

Mizuta, Hiroshi; Uruwashi, Shinichi.

1979-01-01

X-ray cinematography has been used for monitoring fuel motion in the out-of-pile fuel pin joule melting experiments for nuclear, liquid metal cooled fast breeder reactor, safety studies related to fuel pin failure, initial fuel motion and thermal fuel-coolant interaction (FCI) of the hypothetical core distractive accident. In order to visually observe the nuclear fuel motion, the X-ray cinematography system consists of an X-ray source located about 5 cm from the test section and an image intensifier located at a corresponding position on the opposite side of the test section. The image from the image intensifier has been recorded both with a high speed camera and video recorder. (author)

Nuclear fuel element

International Nuclear Information System (INIS)

Penrose, R.T.; Thompson, J.R.

1976-01-01

A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

Characteristics of spray from a GDI fuel injector for naphtha and surrogate fuels

KAUST Repository

Wang, Libing

2016-11-18

Characterization of the spray angle, penetration, and droplet size distribution is important to analyze the spray and atomization quality. In this paper, the spray structure development and atomization characterization of two naphtha fuels, namely light naphtha (LN) and whole naphtha (WN) and two reference fuel surrogates, i.e. toluene primary reference fuel (TPRF) and primary reference fuel (PRF) were investigated using a gasoline direct injection (GDI) fuel injector. The experimental setup included a fuel injection system, a high-speed imaging system, and a droplet size measurement system. Spray images were taken by using a high-speed camera for spray angle and penetration analysis. Sauter mean diameter, Dv(10), Dv(50), Dv(90), and particle size distribution were measured using a laser diffraction technique. Results show that the injection process is very consistent for different runs and the time averaged spray angles during the measuring period are 103.45°, 102.84°, 102.46° and 107.61° for LN, WN, TPRF and PRF, respectively. The spray front remains relatively flat during the early stage of the fuel injection process. The peak penetration velocities are 80 m/s, 75 m/s, 75 m/s and 79 m/s for LN, WN, TPRF and PRF, respectively. Then velocities decrease until the end of the injection and stay relatively stable. The transient particle size and the time-averaged particle size were also analyzed and discussed. The concentration weighted average value generally shows higher values than the arithmetic average results. The average data for WN is usually the second smallest except for Dv90, of which WN is the biggest. Generally the arithmetic average particle sizes of PRF are usually the smallest, and the sizes does not change much with the measuring locations. For droplet size distribution results, LN and WN show bimodal distributions for all the locations while TPRF and PRF shows both bimodal and single peak distribution patterns. The results imply that droplet size

LMFBR fuel analysis. Task A: Oxide fuel dynamics. Final report, October 1, 1976--September 30, 1977

International Nuclear Information System (INIS)

Dhir, V.K.; Doshi, J.; Frank, M.; Hauss, B.; Kastenberg, W.E.; Wong, K.

1977-10-01

The study presented deals with several areas of uncertainty in the analysis of the unprotected overpower transient for the Clinch River Breeder Reactor. These areas of uncertainty include the time, place, and mode of fuel pin failure; pre-failure fuel motion; fuel freezing, plugging, and plate-out following pin failure; and the potential for re-criticality. Internal molten fuel motion prior to pin failure was found to be sensitive to ramp rate and burnup. The strain-limit fuel failure criterion was found to be inappropriate for analysis based on existing data. The coupling of pre-transient- and transient-induced stresses tended to force the failure location towards the core midplane

The interpretation of fuel centre temperature measurements on a suspected leaking fuel pin

International Nuclear Information System (INIS)

Ainscough, J.B.; Lang, C.; Clough, D.J.

1983-01-01

In order to study fuel densification a series of single instrumented pin irradiations has been carried out in the High Pressure Water Loop of DIDO at Harwell. The behaviour of two of these pins was different from that expected. In the fifth test, where the fuel was 95% dense pellet UO 2 and expected to densify readily in-reactor, the fuel centre temperature increased from its starting value of approx. 1300 deg. C at a rate somewhat higher than expected on the basis of predicted densification rates. After about six days, the temperature increased rapidly and unexpectedly to 2100-2200 deg. C and remained steady at this level for a further eight days until a reactor trip occurred and the pin was unloaded. Predictions made using the HOTROD code imply a maximum fuel temperature of less than 1500 deg. C after densification. Post-irradiation examination confirmed that fission gas release had occurred, that the measured temperatures were consistent with the fuel microstructure and that the pin had a high internal gas pressure. The fourth pin in the series contained 97% dense UO 2 which was also expected to be dimensionally unstable. Qualitatively its behaviour was similar to that of the fifth pin though the temperatures throughout were lower. This pin experienced a number of major power cycles and failed after about 30 days in-reactor. It is probable that coolant ingress occurred in both pins via the thermocouple Hoke seal, degrading the filling gas conductivity and allowing the fuel to densify rapidly with consequent increase in the fuel/clad gap and hence in fuel temperature. These irradiations show that, for a short time at least, an apparently unfailed pin could operate undetected with temperatures significantly higher than those predicted for normal operation. (author)

Failing Decision

DEFF Research Database (Denmark)

Knudsen, Morten

2014-01-01

Recently the Danish subway trains have begun to announce “on time” when they arrive at a station on time. This action reflects a worrying acceptance of the normality of failure. If trains were generally expected to be on time, there would be no reason to – triumphantly – announce it. This chapter...... by an interest in failure as one way of improving understanding of present-day decision making in organizations.......Recently the Danish subway trains have begun to announce “on time” when they arrive at a station on time. This action reflects a worrying acceptance of the normality of failure. If trains were generally expected to be on time, there would be no reason to – triumphantly – announce it. This chapter...... deals not with traffic delays, but with failing decisions in organizations. The assumption of this chapter is that failing decisions today are as normal as delayed trains. Instead of being the exception, failure is part of the everyday reproduction of organizations – as an uncontrolled effect but also...

Simulating fail-stop in asynchronous distributed systems

Science.gov (United States)

Sabel, Laura; Marzullo, Keith

1994-01-01

The fail-stop failure model appears frequently in the distributed systems literature. However, in an asynchronous distributed system, the fail-stop model cannot be implemented. In particular, it is impossible to reliably detect crash failures in an asynchronous system. In this paper, we show that it is possible to specify and implement a failure model that is indistinguishable from the fail-stop model from the point of view of any process within an asynchronous system. We give necessary conditions for a failure model to be indistinguishable from the fail-stop model, and derive lower bounds on the amount of process replication needed to implement such a failure model. We present a simple one-round protocol for implementing one such failure model, which we call simulated fail-stop.

Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

International Nuclear Information System (INIS)

Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

2017-01-01

Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

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

Fuel quality issues in stationary fuel cell systems.

Energy Technology Data Exchange (ETDEWEB)

Papadias, D.; Ahmed, S.; Kumar, R. (Chemical Sciences and Engineering Division)

2012-02-07

Fuel cell systems are being deployed in stationary applications for the generation of electricity, heat, and hydrogen. These systems use a variety of fuel cell types, ranging from the low temperature polymer electrolyte fuel cell (PEFC) to the high temperature solid oxide fuel cell (SOFC). Depending on the application and location, these systems are being designed to operate on reformate or syngas produced from various fuels that include natural gas, biogas, coal gas, etc. All of these fuels contain species that can potentially damage the fuel cell anode or other unit operations and processes that precede the fuel cell stack. These detrimental effects include loss in performance or durability, and attenuating these effects requires additional components to reduce the impurity concentrations to tolerable levels, if not eliminate the impurity entirely. These impurity management components increase the complexity of the fuel cell system, and they add to the system's capital and operating costs (such as regeneration, replacement and disposal of spent material and maintenance). This project reviewed the public domain information available on the impurities encountered in stationary fuel cell systems, and the effects of the impurities on the fuel cells. A database has been set up that classifies the impurities, especially in renewable fuels, such as landfill gas and anaerobic digester gas. It documents the known deleterious effects on fuel cells, and the maximum allowable concentrations of select impurities suggested by manufacturers and researchers. The literature review helped to identify the impurity removal strategies that are available, and their effectiveness, capacity, and cost. A generic model of a stationary fuel-cell based power plant operating on digester and landfill gas has been developed; it includes a gas processing unit, followed by a fuel cell system. The model includes the key impurity removal steps to enable predictions of impurity breakthrough

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

International Nuclear Information System (INIS)

Boczar, P.G.; Van Dyk, M.T.

1987-02-01

A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

Wet motor gerotor fuel pump with self-aligning bearing

Energy Technology Data Exchange (ETDEWEB)

Carleton, W.A.

1987-02-24

A wet motor gerotor fuel pump is described for pumping fuel from a fuel source to an internal combustion engine comprising: a pump case having one end, an opposite end and a flow axis therethrough, the pump case further comprising an inlet end bore at the one end adapted to communicate with the fuel source; an inlet chamber adjacent to the inlet end bore; a motor chamber located in the opposite end of the pump case; a pump chamber interposed the motor chamber and the inlet chamber; first means for sealing the pump case, the first means for sealing located at the opposite end of the pump case; inlet housing means mounted in the pump chamber, the inlet housing means comprising an annular hub protruding into the inlet chamber. The inlet housing means further comprises a gerotor cavity about a gerotor axis located parallel to and displaced a predetermined distance in an eccentric radial direction from the flow axis.

Large-scale fuel cycle centres

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

Powder handling for automated fuel processing

International Nuclear Information System (INIS)

Frederickson, J.R.; Eschenbaum, R.C.; Goldmann, L.H.

1989-01-01

Installation of the Secure Automated Fabrication (SAF) line has been completed. It is located in the Fuel Cycle Plant (FCP) at the Department of Energy's (DOE) Hanford site near Richland, Washington. The SAF line was designed to fabricate advanced reactor fuel pellets and assemble fuel pins by automated, remote operation. This paper describes powder handling equipment and techniques utilized for automated powder processing and powder conditioning systems in this line. 9 figs

Fuel starvation. Irreversible degradation mechanisms in PEM fuel cells

Energy Technology Data Exchange (ETDEWEB)

Rangel, Carmen M.; Silva, R.A.; Travassos, M.A.; Paiva, T.I.; Fernandes, V.R. [LNEG, National Laboratory for Energy and Geology, Lisboa (Portugal). UPCH Fuel Cells and Hydrogen Unit

2010-07-01

PEM fuel cell operates under very aggressive conditions in both anode and cathode. Failure modes and mechanism in PEM fuel cells include those related to thermal, chemical or mechanical issues that may constrain stability, power and lifetime. In this work, the case of fuel starvation is examined. The anode potential may rise to levels compatible with the oxidization of water. If water is not available, oxidation of the carbon support will accelerate catalyst sintering. Diagnostics methods used for in-situ and ex-situ analysis of PEM fuel cells are selected in order to better categorize irreversible changes of the cell. Electrochemical Impedance Spectroscopy (EIS) is found instrumental in the identification of fuel cell flooding conditions and membrane dehydration associated to mass transport limitations / reactant starvation and protonic conductivity decrease, respectively. Furthermore, it indicates that water electrolysis might happen at the anode. Cross sections of the membrane catalyst and gas diffusion layers examined by scanning electron microscopy indicate electrode thickness reduction as a result of reactions taking place during hydrogen starvation. Catalyst particles are found to migrate outwards and located on carbon backings. Membrane degradation in fuel cell environment is analyzed in terms of the mechanism for fluoride release which is considered an early predictor of membrane degradation. (orig.)

Study on the behavior of unirradiated light water reactor fuel with iodine-127 under the reactivity initiated accident (RIA) conditions

International Nuclear Information System (INIS)

Sasajima, Hideo; Yanagisawa, Kazuaki; Kanazawa, Hiroyuki

1988-07-01

In a light water reactor fuel, a stress arised from pellet-cladding interaction (PCI) will have possibility to cause stress corrosion claddling (PCI failure) under an environment of corrosive fission product ; line iodine. A pulse irradiation experiment by NSRR was carried out to examine whether or not an unirradiated NSRR standard fuel rod in which 1.3 mg (33 x 10 -6 g/cm 2 ) of iodine was artificially filled could cause the PCI failure. Obtained results are: (1) The fuel rods with iodine did not fail both at deposited energy levels of 268 and 280 cal/g · UO 2 . On the other hand, the fuel rods without iodine failed at the same energy levels due to thinning of the cladding wall thickness. Within this experimental scope, PCI-failure did not occur on iodine filled fuel rods. (2) At a periphery of the fuel pellet of iodine filled rod, an uniform torus ring was formed. The torus ring consisted of an equi-axed large grains at 268 cal/g · UO 2 and a columnar ones at 280 cal/g · UO 2 . The torus ring was not formed in the fuel without iodine. (author)

Improved Retrieval Technique of pin-wise composition for spent fuel recycling

Energy Technology Data Exchange (ETDEWEB)

Park, YunSeo; Kim, Myung Hyun [Kyung Hee University , Yongin (Korea, Republic of)

2016-10-15

New reutilization method which does not require fabrication processing was suggested and showed feasibility by Dr. Aung Tharn Daing. This new reutilization method is predict spent nuclear fuel pin composition, reconstruct new fuel assembly by spent nuclear pin, and directly reutilize in same PWR core. There are some limitation to predict spent nuclear fuel pin composition on his methodology such as spatial effect was not considered enough. This research suggests improving Dr. Aung Tharn Daing's retrieval technique of pin-wise composition. This new method classify fuel pin groups by its location effect in fuel assembly. Most of fuel pin composition along to burnup in fuel assembly is not highly dependent on location. However, compositions of few fuel pins where near water hole and corner of fuel assembly are quite different in same burnup. Required number of nuclide table is slightly increased from 3 to 6 for one fuel assembly with this new method. Despite of this little change, prediction of the pin-wise composition became more accurate. This new method guarantees two advantages than previous retrieving technique. First, accurate pin-wise isotope prediction is possible by considering location effect in a fuel assembly. Second, it requires much less nuclide tables than using full single assembly database. Retrieving technique of pin-wise composition can be applied on spent fuel management field useful. This technique can be used on direct use of spent fuel such as Dr. Aung Tharn Daing showed or applied on pin-wise waste management instead of conventional assembly-wise waste management.

Surveillance instrumentation for spent-fuel safeguards

International Nuclear Information System (INIS)

McKenzie, J.M.; Holmes, J.P.; Gillman, L.K.; Schmitz, J.A.; McDaniel, P.J.

1978-01-01

The movement, in a facility, of spent reactor fuel may be tracked using simple instrumentation together with a real time unfolding algorithm. Experimental measurements, from multiple radiation monitors and crane weight and position monitors, were obtained during spent fuel movements at the G.E. Morris Spent-Fuel Storage Facility. These data and a preliminary version of an unfolding algorithm were used to estimate the position of the centroid and the magnitude of the spent fuel radiation source. Spatial location was estimated to +-1.5 m and source magnitude to +-10% of their true values. Application of this surveillance instrumentation to spent-fuel safeguards is discussed

Electronuclear fissile fuel production. Linear accelerator fuel regenerator and producer LAFR and LAFP

International Nuclear Information System (INIS)

Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

1978-04-01

A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for 3 to 1000 MW(e) LWR power reactors over its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center

Reliability analysis of dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Reliability analysis of dispersion nuclear fuel elements

Energy Technology Data Exchange (ETDEWEB)

Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

2008-03-15

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Power ramping test in the JMTR for PCI study of water reactor fuel

International Nuclear Information System (INIS)

Nakata, H.; Kanbara, M.; Ichikawa, M.

1984-01-01

Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

Nuclear fuel storage

International Nuclear Information System (INIS)

Bevilacqua, F.

1981-01-01

A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

Safeguards instrument to monitor spent reactor fuel

International Nuclear Information System (INIS)

Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.

1981-01-01

A hand-held instrument for monitoring irradiated nuclear fuel inventories located in water-filled storage ponds has been developed. This instrument provides sufficient precise qualitative and quantitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors, and is believed to be of potential use to nuclear fuel managers and to operators of spent-fuel storage facilities, both at reactor and away-from-reactor, and to operators of nuclear fuel reprocessing plants. Because the Cerenkov radiation glow can barely be seen by the unaided eye under darkened conditions, a night vision device is incorporated to aid the operator in locating the fuel assembly to be measured. Beam splitting optics placed in front of the image intensifier and a preset aperture select a predetermined portion of the observed scene for measurement of the light intensity using a photomultiplier (PM) tube and digital readout. The PM tube gain is adjusted by use of an internal optical reference source, providing long term repeatability and instrument-to-instrument cnsistency. Interchangeable lenses accommodate various viewing and measuring conditions

Financial subsidies to the Australian fossil fuel industry

International Nuclear Information System (INIS)

Riedy, Chris; Diesendorf, Mark

2003-01-01

A common claim during international greenhouse gas reduction negotiations has been that domestic emissions cuts will harm national economies. This argument fails to consider the distorting effect of existing financial subsidies and associated incentives to fossil fuel production and consumption provided by governments in most developed countries. These subsidies support a fossil fuel energy sector that is the major contributor to global greenhouse gas emissions and conflict with attempts to expand the role of sustainable energy technologies. Reform of these types of subsidies has the potential to provide substantial gains in economic efficiency as well as reductions in carbon dioxide emissions--a 'no regrets' outcome for the economy and the environment. This paper examines financial subsidies to fossil fuel production and consumption in Australia and estimates the magnitude of the subsidies. Subsidies and associated incentives to fossil fuel production and consumption in Australia are similar to those in the United States and the other countries that have pushed for increased 'flexibility' during international negotiations

Spent fuel surveillance and monitoring methods

International Nuclear Information System (INIS)

1988-05-01

The Technical Committee Meeting on ''Spent Fuel Surveillance and Monitoring Methods'' (27-30 October 1987) has been organized in accordance with recommendations of the International Standing Advisory Group on Spent Fuel Management during its second meeting in 1986. The aim of the meeting was to discuss the above questions with emphasis on current design and operation criteria, safety principles and licensing requirements and procedures in order to prevent: inadvertent criticality, undue radiation exposure, unacceptable release of radioactivity as well as control for loss of storage pool water, crud impact, water chemistry, distribution and behaviour of particulates in cooling water, oxidation of intact and failed fuel rods as a function of temperature and burnup; distribution of radiation and temperature through dry cask wall, monitoring of leakages from pools and gas escapes from dry storage facilities, periodical integrity tests of the containment barriers, responsibilities of organizations for the required operation, structure, staff and subordination, etc. The presentations of the Meeting were divided into two sessions: Spent fuel surveillance programmes and practice in Member States (4 papers); Experimental methods developed in support of spent fuel surveillance programmes (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Improvement of biogas as a fuel in molten carbonate fuel cells (MCFC); Aprovechamiento del biogas como combustible en pilas de combustible de carbonatos fundidos (MCFC)

Energy Technology Data Exchange (ETDEWEB)

Gil Diez, J.

2002-07-01

Molten carbonate fuel cells (MCFC) have a high efficiency of approx 50% when using biogas as a fuel and are among all types of FC the best suited for biogas. A precondition for use biogas in fuel cells is the reduction of accompanying traces of detrimental gases, therefore the RTD-work is two fold: A Preprocessing unit must be developed and the expected endurance must be confirmed. As a lesson learned in prior projects major reasons why renewable energy projects fail is the one-sided focus on technical aspects, that is why non-technical barriers shall be taken into account and realistic recommendations have to be established to overcome possible economic, logistic, legal and social problems. (Author)

Capabilities for processing shipping casks at spent fuel storage facilities

International Nuclear Information System (INIS)

Baker, W.H.; Arnett, L.M.

1978-01-01

Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

Storage of spent nuclear fuel

International Nuclear Information System (INIS)

Machado, O.J.; Moore, J.T.; Cooney, B.F.

1989-01-01

This patent describes a rack for storing nuclear fuel assemblies. The rack including a base, an array of side-by-side fuel-storage locations, each location being a hollow body of rectangular transverse cross section formed of metallic sheet means which is readily bent, each body having a volume therein dimensioned to receive a fuel assembly. The bodies being mounted on the base with each body secured to bodies adjacent each body along welded joints, each joint joining directly the respective contiguous corners of each body and of bodies adjacent to each body and being formed by a series of separate welds spaced longitudinally between the tops and bottoms of the secured bodies along each joint. The spacings of the separate welds being such that the response of the rack when it is subjected to the anticipated seismic acceleration of the rack, characteristic of the geographical regions where the rack is installed, is minimized

Method of detecting a fuel element failure

International Nuclear Information System (INIS)

Cohen, P.

1975-01-01

A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

Utility residual fuel oil market conditions: An update

International Nuclear Information System (INIS)

Mueller, H.A. Jr.

1992-01-01

Planning for residual fuel oil usage and management remains an important part of the generation fuel planning and management function for many utilities. EPRI's Utility Planning Methods Center has maintained its analytical overview of the fuel oil markets as part of its overall fuel planning and management research program. This overview provides an update of recent fuel oil market directions. Several key events of the past year have had important implications for residual fuel oil markets. The key events have been the changes brought about by the Persian Gulf War and its aftermath, as well as continuing environmental policy developments. The Persian Gulf conflict has created renewed interest in reducing fuel oil use by utilities as part of an overall reduction in oil imports. The policy analysis performed to date has generally failed to properly evaluate utility industry capability. The Persian Gulf conflict has also resulted in an important change in the structure of international oil markets. The result of this policy-based change is likely to be a shift in oil pricing strategy. Finally, continued change in environmental requirements is continuing to shift utility residual oil requirements, but is also changing the nature of the US resid market itself

Why did occidental modernity fail in the Arab Middle East: the failed modern state?

OpenAIRE

Sardar, Aziz

2011-01-01

This thesis asks a straightforward but nevertheless a complex question, that is: Why did modernity fail in the Arab Middle East? The notion of modernity in this thesis signifies the occidental modernity which reached the region in many different forms and through various channels. This occidental modernity had an impact on many areas and changed the societies and politics of the region. But these changes stopped short of reaching modernity, in other words it failed to change the society from ...

Nuclear reactor fuel element assemblies

International Nuclear Information System (INIS)

Raven, L.F.

1975-01-01

A spacer grid for a nuclear fuel element comprises a plurality of cojointed cylindrical ferrules adapted to receive a nuclear fuel pin. Each ferrule has a pair of circumferentially spaced rigid stop members extending inside the ferrule and a spring locating member attached to the ferrule and also extending from the ferrule wall inwardly thereof at such a circumferential spacing relative to the rigid stop members that the line of action of the spring locating member passes in opposition to and between the rigid stop members which lie in the same diametric plane. At least some of the cylindrical ferrules have one rim shaped to promote turbulence in fluid flowing through the grid. (Official Gazette)

HTGR fuel behavior at very high temperature

International Nuclear Information System (INIS)

Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

1986-03-01

Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

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

Nuclear instrumentation systems in prototype fast breeder reactor

International Nuclear Information System (INIS)

Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

2004-01-01

The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

Combustion of fuels with low sintering temperature

Energy Technology Data Exchange (ETDEWEB)

Dalin, D

1950-08-16

A furnace for the combustion of low sintering temperature fuel consists of a vertical fuel shaft arranged to be charged from above and supplied with combustion air from below and containing a system of tube coils extending through the fuel bed and serving the circulation of a heat-absorbing fluid, such as water or steam. The tube-coil system has portions of different heat-absorbing capacity which are so related to the intensity of combustion in the zones of the fuel shaft in which they are located as to keep all parts of the fuel charge below sintering temperature.

Analysis of possibilities for functional capacity for work rise of reactor fuel elements at nuclear engine regime

International Nuclear Information System (INIS)

Deryavko, I.I.; Perepelkin, I.G.; Pivovarov, O.S.; Storozhenko, A.N.; Tarasov, V.I.

2000-01-01

The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

Evaluating alternative fuel treatment strategies to reduce wildfire losses in a Mediterranean area

Science.gov (United States)

Michele Salis; Maurizio Laconi; Alan A. Ager; Fermin J. Alcasena; Bachisio Arca; Olga Lozano; Ana Fernandes de Oliveira; Donatella Spano

2016-01-01

The goal of this work is to evaluate by a modeling approach the effectiveness of alternative fuel treatment strategies to reduce potential losses from wildfires in Mediterranean areas. We compared strategic fuel treatments located near specific human values vs random locations, and treated 3, 9 and 15% of a 68,000 ha study area located in Sardinia, Italy. The...

Expandable device for a nuclear fuel rod

International Nuclear Information System (INIS)

Gesinski, L.T.

1978-01-01

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

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

Fuel-motion diagnostics and cineradiography

International Nuclear Information System (INIS)

DeVolpi, A.

1982-09-01

Nuclear and non-nuclear applications of cineradiography are reviewed, with emphasis on diagnostic instrumentation for in-pile transient-reactor safety testing of nuclear fuel motion. The primary instrument for this purpose has been the fast-neutron hodoscope, which has achieved quantitative monitoring of time, location, mass, and velocity of fuel movement under the difficult conditions associated with transient-reactor experiments. Alternative diagnostic devices that have been developed have not matched the capabilities of the hodoscope. Other applications for the fuel-motion diagnostic apparatus are also evolving, including time-integrated radiography and direct time- and space-resolved fuel-pin power monitoring. Although only two reactors are now actively equipped with high-resolution fuel-motion diagnostic systems, studies and tests have been carried out in and for many other reactors

Large-scale fuel cycle centers

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

WWER expert system for fuel failure analysis using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

2008-01-01

The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

1997-06-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

California's Low-Carbon Fuel Standard - Compliance Trends

Science.gov (United States)

Witcover, J.; Yeh, S.

2013-12-01

Policies to incentivize lower carbon transport fuels have become more prevalent even as they spark heated debate over their cost and feasibility. California's approach - performance-based regulation called the Low Carbon Fuel Standard (LCFS) - has proved no exception. The LCFS aims to achieve 10% reductions in state transport fuel carbon intensity (CI) by 2020, by setting declining annual CI targets, and rewarding fuels for incremental improvements in CI beyond the targets while penalizing those that fail to meet requirements. Even as debate continues over when new, lower carbon fuels will become widely available at commercial scale, California's transport energy mix is shifting in gradual but noticeable ways under the LCFS. We analyze the changes using available data on LCFS fuels from the California Air Resources Board and other secondary sources, beginning in 2011 (the first compliance year). We examine trends in program compliance (evaluated through carbon credits and deficits generated), and relative importance of various transport energy pathways (fuel types and feedstocks, and their CI ratings, including new pathways added since the program's start). We document a roughly 2% decline in CI for gasoline and diesel substitutes under the program, with compliance achieved through small shifts toward greater reliance on fuels with lower CI ratings within a relatively stable amount of transport energy derived from alternatives to fossil fuel gasoline and diesel. We also discuss price trends in the nascent LCFS credit market. The results are important to the broader policy debate about transportation sector response to market-based policies aimed at reducing the sector's greenhouse gas emissions.

Closed External Fixation for Failing or Failed Femoral Shaft Plating in a Developing Country.

Science.gov (United States)

Aliakbar, Adil; Witwit, Ibrahim; Al-Algawy, Alaa A Hussein

2017-08-01

Femoral shaft fractures are one of the common injuries that is treated by open reduction, with internal fixation by plate and screws or intramedullary nailing, which can achieve a high union rate. To evaluate the outcome of using closed external fixation to augment a failing plate; with signs of screw loosening and increasing bone/plate gap; a failed plate; broken plate; screws completely out of bone with redisplacement of fracture. A retrospective study on 18 patients, aged between 17-42 years, who presented between 6-18 weeks after initial surgical fixation, with pain, difficulty in limb function, deformity and abnormal movement at fracture site, was done. X-Rays showed plating failure with acceptable amount of callus, which unfortunately had refractured. Cases associated with infection and no radiological evidence of callus formation were excluded from this study. Closed reduction was done by manipulation, then fracture fixation by AO external fixator. The patients were encouraged for full weight bearing as early as possible with dynamization later on. Of the 18 patients who underwent external fixation after close reduction, 15 cases showed bone healing in a period between 11-18 weeks (mean of 14.27 weeks) with good alignment (Radiologically). Removal of external fixator was done followed by physical therapy thereafter. Closed external fixation for treatment of failing or failed femoral plating, achieves good success rate and has less complications, is a short time procedure, especially in a hospital with limited resources.

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

International Nuclear Information System (INIS)

Thoms, K.R.

1975-04-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 (Ta-8 percent W-2 percent Hf) and uranium dioxide (UO 2 ) fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1 percent Zr. A total of nine fuel pins was irradiated (four containing porous UN, two containing dense, nonporous UN, and three containing dense UO 2 ) at average cladding temperatures ranging from 931 to 1015 0 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO 2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of 10 -5 . (U.S.)

30 CFR 75.1905-1 - Diesel fuel piping systems.

Science.gov (United States)

2010-07-01

... facility. (g) Diesel fuel piping systems from the surface shall only be used to transport diesel fuel... storage facility. (h) The diesel fuel piping system must not be located in a borehole with electric power... entry as electric cables or power lines. Where it is necessary for piping systems to cross electric...

Architecture Study for a Fuel Depot Supplied from Lunar Assets

Science.gov (United States)

Perrin, Thomas M.; Casler, James G.

2016-01-01

This architecture study sought to determine the optimum architecture for a fuel depot supplied from lunar assets. Four factors - the location of propellant processing (on the Moon or on the depot), the depot location (on the Moon, L1, GEO, or LEO), the propellant transfer location (L1, GEO, or LEO), and the propellant transfer method (bulk fuel or canister exchange) were combined to identify 18 candidate architectures. Two design reference missions (DRMs) - a commercial satellite servicing mission and a Government cargo mission to Mars - created demand for propellants, while a propellant delivery DRM examined supply issues. The study concluded Earth-Moon L1 is the best location for an orbiting depot. For all architectures, propellant boiloff was less than anticipated, and was far overshadowed by delta-v requirements and resulting fuel consumption. Bulk transfer is the most flexible for both the supplier and customer. However, since canister exchange bypasses the transfer of bulk cryogens and necessary chilldown losses, canister exchange shows promise and merits further investigation. Overall, this work indicates propellant consumption and loss is an essential factor in assessing fuel depot architectures.

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

International Nuclear Information System (INIS)

Yancey, M.E.

1979-01-01

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

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

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

1984-01-01

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

Neglected City Narratives And Failed Rebranding

DEFF Research Database (Denmark)

Mousten, Birthe; Locmele, Gunta

2017-01-01

Rīga, Latvia went through a failed rebranding process as the forerunner of its status as a European Capital of Culture (2014). The same thing happened in Aarhus, Denmark. Aarhus will be a European Capital of Culture (2017) and leading to this, it went through a failed rebranding process. Based on...

When Organization Fails: Why Authority Matters

DEFF Research Database (Denmark)

Blaschke, Steffen

2015-01-01

Review of: James R. Taylor and Elizabeth J. Van Every / When Organization Fails: Why Authority Matters. (New York: Routledge, 2014. 220 pp. ISBN: 978 0415741668)......Review of: James R. Taylor and Elizabeth J. Van Every / When Organization Fails: Why Authority Matters. (New York: Routledge, 2014. 220 pp. ISBN: 978 0415741668)...

Fuel cell water transport

Science.gov (United States)

Vanderborgh, Nicholas E.; Hedstrom, James C.

1990-01-01

The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

The cost of fuel economy in the Indian passenger vehicle market

International Nuclear Information System (INIS)

Chugh, Randy; Cropper, Maureen; Narain, Urvashi

2011-01-01

To investigate how fuel economy is valued in the Indian car market, we compute the cost to Indian consumers of purchasing a more fuel-efficient vehicle and compare it to the benefit of lower fuel costs over the life of the vehicle. We estimate hedonic price functions for four market segments (petrol hatchbacks, diesel hatchbacks, petrol sedans, and diesel sedans) to compute 95% confidence intervals for the marginal cost to the consumer for an increase in fuel economy. We find that the associated present value of fuel savings falls within the 95% confidence interval for most specifications, in all market segments, for the years 2002 through 2006. Thus, we fail to consistently reject the hypothesis that consumers appropriately value fuel economy. - Highlights: → We examine the tradeoffs faced by new vehicle consumers in India. → We use hedonic price functions and instrumental variables. → We find no support for the hypothesis that consumers undervalue fuel economy. → Some consumers are willing to forgo substantial potential savings to own their preferred vehicle.

Strength of Anode‐Supported Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Faes, A.; Frandsen, Henrik Lund; Kaiser, Andreas

2011-01-01

Nickel oxide and yttria doped zirconia composite strength is crucial for anode‐supported solid oxide fuel cells, especially during transient operation, but also for the initial stacking process, where cell curvature after sintering can cause problems. This work first compares tensile and ball....... Even though the electrolyte is to the tensile side, it is found that the anode support fails due to the thermo‐mechanical residual stresses....

On-site concrete cask storage system for spent nuclear fuel

International Nuclear Information System (INIS)

Craig, P.A.; Haelsig, R.T.; Kent, J.D.; Schmoker, D.S.

1989-01-01

A method is described of storing spent nuclear fuel assemblies including the steps of: transferring the fuel assemblies from a spent-fuel pool to a moveable concrete storage cask located outside the spent-fuel pool; maintaining a barrier between the fuel and the concrete in the cask to prevent contamination of the concrete by the fuel; maintaining the concrete storage cask containing the spent-fuel on site at the reactor complex for some predetermined period; transferring the fuel assemblies from the concrete storage cask to a shipping container; and, recycling the concrete storage cask

7 CFR 983.52 - Failed lots/rework procedure.

Science.gov (United States)

2010-01-01

... 7 Agriculture 8 2010-01-01 2010-01-01 false Failed lots/rework procedure. 983.52 Section 983.52..., ARIZONA, AND NEW MEXICO Regulations § 983.52 Failed lots/rework procedure. (a) Substandard pistachios... committee may establish, with the Secretary's approval, appropriate rework procedures. (b) Failed lot...

K Basins fuel encapsulation and storage hazard categorization

International Nuclear Information System (INIS)

Porten, D.R.

1994-12-01

This document establishes the initial hazard categorization for K-Basin fuel encapsulation and storage in the 100 K Area of the Hanford site. The Hazard Categorization for K-Basins addresses the potential for release of radioactive and non-radioactive hazardous material located in the K-Basins and their supporting facilities. The Hazard Categorization covers the hazards associated with normal K-Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. The criteria categorizes a facility based on total curies per radionuclide located in the facility. Tables 5-3 and 5-4 display the results in section 5.0. In accordance with DOE-STD-1027 and the analysis provided in section 5.0, the K East Basin fuel encapsulation and storage activity and the K West Basin storage are classified as a open-quotes Category 2close quotes Facility

Monte-Carlo simulation of dispersion fuel meat structure

International Nuclear Information System (INIS)

Xing Zhonghu; Ying Shihao

2003-01-01

Under the irradiation conditions in research reactors, the inter-diffusion occurs at the fuel particle and matrix interfaces of U 3 Si 2 -Al dispersion fuel. Because of the inter-diffusion reaction, the U 3 Al 7 Si 2 layer is formed around each U 3 Si 2 particle. The layer thickness grows up with irradiation duration and fission density. The formation of resultant layer causes the consumption of U 3 Si 2 fuel and aluminum matrix. This process leads to the evolution of geometrical structure of fuel meat. According to the stochastic locations of particles in dispersion, the authors developed a simulation method for the evolution of the fuel meat structure by utilizing Monte-Carlo method. Every particle is characterized by its diameter and location. The parameters of meat structure include particle size distribution, as-fabricated fuel volume fraction, resultant layer thickness, layer volume fraction, U 3 Si 2 fuel volume fraction, aluminum volume fraction, contiguity probability and inter-linkage fraction of particles. Particularly for the dispersion with as-fabricated fuel volume fraction of 43% and particle sizes in a well-defined normal distribution, more than 13000 sampling particles are simulated in the meat volume of 6 mm x 6 mm x 0.5 mm. The meat structure parameters are calculated as functions of layer thickness in the range from 0-16 μm. (authors)

A geographical analysis of the Swedish wood fuel market

International Nuclear Information System (INIS)

Roos, Anders; Bohlin, Folke; Hektor, Bo; Hillring, Bengt; Parikka, Matti

2000-01-01

The geographical variation in Swedish wood fuel market characteristics for the district heating sector has been studied using Geographical Information Systems (GIS) and cross-sectional Tobit analysis. The results indicate that local availability and competition for wood fuels influence the wood fuel consumption at inland heating plants. The factors affecting the decision to use wood fuel at heating plants close to seaports, however, were not captured by the model, suggesting that coastal location reduces dependency on the local wood fuel market. The effects of changes in local wood fuel availability on wood fuel use by an inland heating plant are presented and discussed

Acceptance of non-fuel assembly hardware by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

Energy Technology Data Exchange (ETDEWEB)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

2016-10-15

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

International Nuclear Information System (INIS)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

2016-01-01

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

Nuclear reactor fuel element assembly spacer grid and method of making

International Nuclear Information System (INIS)

Chetter, J.

1975-01-01

A cellular fuel element assembly spacer grid is described which provides for resilient bracing of fuel pins in the cells of the grid by bow spring locating members projecting inside the cells of the grid to hold the fuel pins against opposed rigid stops also projecting inside the cells of the grid. The grid comprises two tiers each formed from intersecting strip members defining cells which are penetrated by the fuel pins and arranged parallel to one another but spaced apart. The bow spring locating members extend longitudinally between the two tiers and have end ferrules which are a sliding fit on locating members which extend longitudinally from the facing inner edges of the strip members forming the two tiers. The grid tiers are fabricated individually by heat bonding the intersecting strip members prior to assembling the tiers into the spacer grid. (U.S.)

Peripheral pin alignment system for fuel assemblies

International Nuclear Information System (INIS)

Anthony, A.J.

1981-01-01

An alignment system is provided for nuclear fuel assemblies in a nuclear core. The core support structure of the nuclear reactor includes upwardly pointing alignment pins arranged in a square grid and engage peripheral depressions formed in the lateral periphery of the lower ends of each of the fuel assemblies of the core. In a preferred embodiment, the depressions are located at the corners of the fuel assemblies so that each depression includes one-quarter of a cylindrical void. Accordingly, each fuel assembly is positioned and aligned by one-quarter of four separate alignment pins which engage the fuel assemblies at their lower exterior corners. (author)

Spent nuclear fuel storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2005-01-01

When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

Method for monitoring irradiated nuclear fuel using cerenkov radiation

International Nuclear Information System (INIS)

Caldwell, J.T.; Dowdy, E.J.; Nicholson, N.

1983-01-01

A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the cerenkov light intensity measurement is taken at selected bright spots corresponding to the water-filled interstices of the assembly in the water storage, the waterfilled interstices acting as cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the cerenkov radiation intensity also is possible using spot photometers pointed at the assembly

DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Miles, K.J.; Hill, D.J.

1986-01-01

The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

Yalina booster subcritical assembly performance with low enriched uranium fuel

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2011-01-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

Yalina booster subcritical assembly performance with low enriched uranium fuel

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

2011-07-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

Mixed fuel strategy for carbon deposition mitigation in solid oxide fuel cells at intermediate temperatures.

Science.gov (United States)

Su, Chao; Chen, Yubo; Wang, Wei; Ran, Ran; Shao, Zongping; Diniz da Costa, João C; Liu, Shaomin

2014-06-17

In this study, we propose and experimentally verified that methane and formic acid mixed fuel can be employed to sustain solid oxide fuel cells (SOFCs) to deliver high power outputs at intermediate temperatures and simultaneously reduce the coke formation over the anode catalyst. In this SOFC system, methane itself was one part of the fuel, but it also played as the carrier gas to deliver the formic acid to reach the anode chamber. On the other hand, the products from the thermal decomposition of formic acid helped to reduce the carbon deposition from methane cracking. In order to clarify the reaction pathways for carbon formation and elimination occurring in the anode chamber during the SOFC operation, O2-TPO and SEM analysis were carried out together with the theoretical calculation. Electrochemical tests demonstrated that stable and high power output at an intermediate temperature range was well-maintained with a peak power density of 1061 mW cm(-2) at 750 °C. With the synergic functions provided by the mixed fuel, the SOFC was running for 3 days without any sign of cell performance decay. In sharp contrast, fuelled by pure methane and tested at similar conditions, the SOFC immediately failed after running for only 30 min due to significant carbon deposition. This work opens a new way for SOFC to conquer the annoying problem of carbon deposition just by properly selecting the fuel components to realize their synergic effects.

Fuel failure in water reactors: Causes and mitigation. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2003-03-01

The objective of this technical meeting (TM) was to review the present knowledge of the causes and mechanisms of fuel failure in water reactors during normal operational conditions. Emphasis has been given to analysis of failure causes and their mitigation by means of design as well as plant and core operation including strategies for operation with failed fuel. Some information on detection techniques (on-line monitoring and diagnostics, flux tilting, sipping techniques, etc) has also been presented. This TM presented also the progress on the above-mentioned subjects since the last meeting held in 1992 (Dimitrovgrad, Russian Federation). The topics covered in the papers were as follows: Experience feedback on fuel reliability (8 papers); Strategies to avoid or mitigate fuel failures (4 papers); Experimental studies on fuel failures and degradation mechanisms (4 papers); Modelling of fuel failure mechanisms (3 papers); Detection and monitoring during operation or outage (4 papers); Modelling and assessment of fuel failures (3 papers)

Reference thorium fuel cycle

International Nuclear Information System (INIS)

Driggers, F.E.

1978-08-01

In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with 238 U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of 238 U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure

The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

International Nuclear Information System (INIS)

Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

1977-01-01

The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

Survival of dental implants placed in sites of previously failed implants.

Science.gov (United States)

Chrcanovic, Bruno R; Kisch, Jenö; Albrektsson, Tomas; Wennerberg, Ann

2017-11-01

To assess the survival of dental implants placed in sites of previously failed implants and to explore the possible factors that might affect the outcome of this reimplantation procedure. Patients that had failed dental implants, which were replaced with the same implant type at the same site, were included. Descriptive statistics were used to describe the patients and implants; survival analysis was also performed. The effect of systemic, environmental, and local factors on the survival of the reoperated implants was evaluated. 175 of 10,096 implants in 98 patients were replaced by another implant at the same location (159, 14, and 2 implants at second, third, and fourth surgeries, respectively). Newly replaced implants were generally of similar diameter but of shorter length compared to the previously placed fixtures. A statistically significant greater percentage of lost implants were placed in sites with low bone quantity. There was a statistically significant difference (P = 0.032) in the survival rates between implants that were inserted for the first time (94%) and implants that replaced the ones lost (73%). There was a statistically higher failure rate of the reoperated implants for patients taking antidepressants and antithrombotic agents. Dental implants replacing failed implants had lower survival rates than the rates reported for the previous attempts of implant placement. It is suggested that a site-specific negative effect may possibly be associated with this phenomenon, as well as the intake of antidepressants and antithrombotic agents. © 2016 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Nuclear fuel manufacturing. Current activities and prospects at INR Pitesti

International Nuclear Information System (INIS)

Horhoianu, Grigore

2001-01-01

Development of the CANDU nuclear fuel is currently conducted world wide onto two principal directions: - increasing the service span of the current type of fuel and improving the efficiency of burnup in reactor; - reducing the costs of fuel manufacturing by improving the design and manufacturing technologies in condition of increasing fuel performance. In parallel, a research program, RAAN, is undergoing, concerning the development of advanced CANDU type fuels (SEU, RU, DUPIC, Th), aiming at reducing the overall costs per fuel cycle. In the INR TRIGA reactor a large number of experimental fuel elements manufactured in INR were irradiated under different conditions specific to the CANDU reactor operation. Post irradiation investigations both destructive and non-destructive were carried out in the hot cells at INR Pitesti. The experimental results were used in order to optimize and evaluate the fuel project, to check the fuel manufacturing technologies as well as to certify the computational codes. The local thermo-mechanical analyses by final element methods, modelling the SCC phenomenon, probabilistic evaluation of performance parameters of the fuel, constitute new directions in the modelling and developing computational code. The developed codes were submitted to a thorough validation process to comply with the quality assurance. The excellent results obtained in INR were confirmed by participation in the FUMEX International Exercises of computer code intercomparison, organized by IAEA Vienna. Progress was also recorded in establishing the behaviour of fuel elements failed during reactor operation and the effect their maintenance in the reactor core could have upon the power reactor operation. A system-expert variant was worked out able for a short term analysis of the decisions referring to removing the failing element at Cernavoda NPP. As advanced CANDU fuel is concerned, until now preliminary variants for a fuel bundle with 43 elements containing slightly

On the problems of the fuel cycles of nuclear fuels

International Nuclear Information System (INIS)

Schmidt-Kuester, W.J.; Wagner, H.F.

1976-01-01

A secured procurement with nuclear energy can be only achieved if a completely closed fuel cycle will be established. In the Federal Republic of Germany efforts are concentrated on the front end as well as on the back end of the fuel cycle. At the front end the main tasks are to secure uranium supply and to establish the necessary enrichment capacity. The German concept for the back end of the fuel cycle will provide for an integrated and co-located system for all necessary facilities including reprocessing, plutonium fuel fabrication, treatment, interim storage and final disposal of the radioactive wastes to be operational in the mid-80's. Responsibilities for establishing this system are shared between government and private industry. Government will provide for final waste disposal, industry will built and operate the other facilities. Another important point for the introduction of nuclear energy is to solve reliably the problems of protection of fissionable material, radioactive waste and nuclear facilities. German government has initiated respective activities and has started appropriate R+D-work. (orig.) [de

Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

1990-05-01

The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

Design of fuel failure detection system for multipurpose reactor GA. Siwabessy

International Nuclear Information System (INIS)

Sujalmo Saiful; Kuntoro Iman; Sato, Mitsugu; Isshiki, Masahiko.

1992-01-01

A fuel failure detection system (FFDS) has been designed for the Reactor GA. Siwabessy. The FFDS is aimed to detect fuel failure by observing delayed neutron released by fission products such as N-17, I-137, Br-87 and Br-88 in the primary cooling system. The delayed neutrons will be detected by using four neutron detectors, type BF-3, which are located inside a Sampling Tank. The detector location has been determined and the location is associated with the transit time from the reactor core outlet to the Sampling Tank, which is approximately 60 seconds. The neutron detection efficiency was calculated by using a computer code named MORSE. The FFDS has the capability to detect as quickly as possible, even a small failure of a fuel element occurring in the reactor core. Therefore the presence of FFDS in a reactor must be considered, in order to prevent further progress if the fuel failure occurs. (author)

Palliative effects of H2 on SOFCs operating with carbon containing fuels

Science.gov (United States)

Reeping, Kyle W.; Bohn, Jessie M.; Walker, Robert A.

2017-12-01

Chlorine can accelerate degradation of solid oxide fuel cell (SOFC) Ni-based anodes operating on carbon containing fuels through several different mechanisms. However, supplementing the fuel with a small percentage of excess molecular hydrogen effectively masks the degradation to the catalytic activity of the Ni and carbon fuel cracking reaction reactions. Experiments described in this work explore the chemistry behind the "palliative" effect of hydrogen on SOFCs operating with chlorine-contaminated, carbon-containing fuels using a suite of independent, complementary techniques. Operando Raman spectroscopy is used to monitor carbon accumulation and, by inference, Ni catalytic activity while electrochemical techniques including electrochemical impedance spectroscopy and voltammetry are used to monitor overall cell performance. Briefly, hydrogen not only completely hides degradation observed with chlorine-contaminated carbon-containing fuels, but also actively removes adsorbed chlorine from the surface of the Ni, allowing for the methane cracking reaction to continue, albeit at a slower rate. When hydrogen is removed from the fuel stream the cell fails immediately due to chlorine occupation of methane/biogas reaction sites.

Nuclear fuel pin controlled failure device

International Nuclear Information System (INIS)

Schlenker, L.D.

1975-01-01

Each fuel pin of a fuel assembly for a water-cooled nuclear reactor is provided with means for rupturing the cladding tube at a predetermined location if an abnormal increase in pressure of the gases present occurs due to a loss-of-coolant accident. Preferably all such rupture means are oriented to minimize the hydraulic resistance to the flow of emergency core coolant such as all rupture means pointing in the same direction. Rupture means may be disposed at different elevations in adjacent fuel pins and, further, fuel pins may be provided with two or more rupture means, one of which is in the upper portion of the fuel pin. Rupture means are mechanical as by providing a locally weakened condition of a controlled nature in the cladding. (U.S.)

Status of Proposed Repository for Latin-American Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Ferrada, J.J.

2004-10-04

This report compiles preliminary information that supports the premise that a repository is needed in Latin America and analyzes the nuclear situation (mainly in Argentina and Brazil) in terms of nuclear capabilities, inventories, and regional spent-fuel repositories. The report is based on several sources and summarizes (1) the nuclear capabilities in Latin America and establishes the framework for the need of a permanent repository, (2) the International Atomic Energy Agency (IAEA) approach for a regional spent-fuel repository and describes the support that international institutions are lending to this issue, (3) the current situation in Argentina in order to analyze the Argentinean willingness to find a location for a deep geological repository, and (4) the issues involved in selecting a location for the repository and identifies a potential location. This report then draws conclusions based on an analysis of this information. The focus of this report is mainly on spent fuel and does not elaborate on other radiological waste sources.

Nuclear fuel handling apparatus

International Nuclear Information System (INIS)

Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

1977-01-01

A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

Dual winch nuclear fuel transfer system providing more reliable fuel transfer during refueling operations

International Nuclear Information System (INIS)

Meuschke, R.E.; Harper, M.J.; Stefko, D.J.

1991-01-01

This paper describes a nuclear power plant having an auxiliary building, a containment building having the wall, a track extending through a transfer tube within the containment wall, and a fuel transfer system for moving fuel assemblies along the track between the auxiliary building side and the containment building side of the containment wall. It comprises: a car having wheels for movement along spaced rails of the track and further having a carrying basket for one or more fuel assemblies; winch means located on the auxiliary building side of the containment wall and above the water level existing over the track during refueling operations to drive the car along the track; first cable means and second cable means extending substantially vertically downward from the winch means to the tack level; first sheave means for directing the first and the second cable means substantially in the horizontal direction along the track; means for securing the first cable means to the car so that winch pulling force on the first cable means drives the car away from the containment building; second sheave means located near the containment end of the transfer tube

Transport of encapsulated nuclear fuels

International Nuclear Information System (INIS)

Broman, Ulrika; Dybeck, Peter; Ekendahl, Ann-Mari

2005-12-01

The transport system for encapsulated fuel is described, including a preliminary drawing of a transport container. In the report, the encapsulation plant is assumed to be located to Oskarshamn, and the repository to Oskarshamn or Forsmark

A stochastic mathematical model to locate field hospitals under disruption uncertainty for large-scale disaster preparedness

Directory of Open Access Journals (Sweden)

Nezir Aydin

2016-03-01

Full Text Available In this study, we consider field hospital location decisions for emergency treatment points in response to large scale disasters. Specifically, we developed a two-stage stochastic model that determines the number and locations of field hospitals and the allocation of injured victims to these field hospitals. Our model considers the locations as well as the failings of the existing public hospitals while deciding on the location of field hospitals that are anticipated to be opened. The model that we developed is a variant of the P-median location model and it integrates capacity restrictions both on field hospitals that are planned to be opened and the disruptions that occur in existing public hospitals. We conducted experiments to demonstrate how the proposed model can be utilized in practice in a real life problem case scenario. Results show the effects of the failings of existing hospitals, the level of failure probability and the capacity of projected field hospitals to deal with the assessment of any given emergency treatment system’s performance. Crucially, it also specifically provides an assessment on the average distance within which a victim needs to be transferred in order to be treated properly and then from this assessment, the proportion of total satisfied demand is then calculated.

Selection and use of a low enriched fuel in high performance research reactors

International Nuclear Information System (INIS)

Cerles, J.M.; Schwartz, J.P.

1978-08-01

A new nuclear fuel composition for research reactors (Osiris, Siloe) is studied using low enriched (E<20%) uranium oxide. Its utilization leads to modifications in the facilities of these experimental reactors: increase of primary coolant flow, modifications in failed element detection system, handling of materials and storage

Nuclear reactor core and fuel element therefor

International Nuclear Information System (INIS)

Fortescue, P.

1986-01-01

This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

Thermal barrier and support for nuclear reactor fuel core

International Nuclear Information System (INIS)

Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

1987-01-01

A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

Parameter Estimation of Spacecraft Fuel Slosh Model

Science.gov (United States)

Gangadharan, Sathya; Sudermann, James; Marlowe, Andrea; Njengam Charles

2004-01-01

Fuel slosh in the upper stages of a spinning spacecraft during launch has been a long standing concern for the success of a space mission. Energy loss through the movement of the liquid fuel in the fuel tank affects the gyroscopic stability of the spacecraft and leads to nutation (wobble) which can cause devastating control issues. The rate at which nutation develops (defined by Nutation Time Constant (NTC can be tedious to calculate and largely inaccurate if done during the early stages of spacecraft design. Pure analytical means of predicting the influence of onboard liquids have generally failed. A strong need exists to identify and model the conditions of resonance between nutation motion and liquid modes and to understand the general characteristics of the liquid motion that causes the problem in spinning spacecraft. A 3-D computerized model of the fuel slosh that accounts for any resonant modes found in the experimental testing will allow for increased accuracy in the overall modeling process. Development of a more accurate model of the fuel slosh currently lies in a more generalized 3-D computerized model incorporating masses, springs and dampers. Parameters describing the model include the inertia tensor of the fuel, spring constants, and damper coefficients. Refinement and understanding the effects of these parameters allow for a more accurate simulation of fuel slosh. The current research will focus on developing models of different complexity and estimating the model parameters that will ultimately provide a more realistic prediction of Nutation Time Constant obtained through simulation.

In-pile study of the reaction between breeder fuel and sodium

International Nuclear Information System (INIS)

Hugot, J.P.

1982-10-01

Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

Dimensional response of CANDU fuel to power changes

Energy Technology Data Exchange (ETDEWEB)

Fehrenbach, P J [Fuel Engineering Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Hastings, I J; Morel, P A; Sage, R D; Smith, A D [Fuel Materials Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada)

1983-06-01

The introduction of CANLUB-coated fuel cladding, modified fuel management schemes, and revisions to the sequence of control rod movements, have eliminated power ramping fuel failures in CANDU power reactors. However, an irradiation program continues at Chalk River Nuclear Laboratories to determine the effect of various design and operating parameters on the dimensional response of UO{sub 2} fuel elements to power changes, over a range of conditions outside those normally experienced by CANDU power reactor fuel. We have investigated the effect of power changes on element diameter for UO{sub 2} fuel with starting densities of 10.6 and 10.8 Mg/m{sup 3} clad in 0.4 mm thick Zircaloy, at burnups from 0 to 100 MW.h/kg U. Element diameter measurements were obtained at power using an In-Reactor Diameter Measuring Rig (IRDMR). Rates of power change over the range 0.0005 to 0.03 kW.m{sup -1}.s{sup -1} were achieved by a combination of reactor power control and use of a Helium-3 power cycling facility. Total diameter increases in unirradiated elements were about 1% at pellet interface locations for both fuel densities during the initial power increase to 60 kW/m. Diameter changes during subsequent power cycles of these elements from 55 to 100% maximum power were significantly larger for the higher density fuel, ranging from 0.3 to 0.5% compared to less than 0.1% for the standard density (10.6 Mg/m{sup 3}) fuel. In elements pre-irradiated at 27 kW/m to burnups of about 100 MW.h/kg U prior to power ramping, the diameter increases measured after ramping to 55 kW/m also varied with starting fuel density. Diameter changes at pellet interface locations were about 0.9% and 0.6% for higher density and standard density fuel respectively. (author)

Pyroprocessing of IFR Metal Fuel

International Nuclear Information System (INIS)

Laidler, J.J.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle features the use of an innovative reprocessing method, known as open-quotes pyroprocessingclose quotes featuring fused-salt electrofining of the spent fuel. Electrofining of IFR spent fuel involves uranium recovery by electro-transport to a solid steel cathode. The thermodynamics of the system preclude plutonium recovery in the same way, so a liquid cadmium cathode located in the electrolyte salt phase is utilized. The deposition of Pu, Am, Np, and Cm takes place at the liquid cadmium cathode in the form of cadmium intermetallic compounds (e.g, PuCd 6 ), and uranium deposits as the pure metal when cadmium saturation is reached. A small amount of rare earth fission products deposit together with the heavy metals at both the solid and liquid cadmium cathodes, providing a significant degree of self-protection. A full scope demonstration of the IFR fuel cycle will begin in 1993, using fuel irradiated in EBR-II

Pass/fail patterns of candidates who failed COMLEX-USA level 2-PE because of misrepresentation of clinical findings on postencounter notes.

Science.gov (United States)

Langenau, Erik E; Sandella, Jeanne M

2011-07-01

In 2007, The National Board of Osteopathic Medical Examiners (NBOME) instituted a policy to address the accuracy and integrity of postencounter written documentation recorded during the Comprehensive Osteopathic Medical Licensing Examination Level 2-Performance Evaluation (COMLEX-USA Level 2-PE). This policy was instituted not only to protect the integrity of the examination, but also to highlight that overdocumentation of clinical findings not obtained during patient encounters may jeopardize patient safety. To investigate overall and domain pass/fail patterns of candidates who misrepresented clinical findings with regard to past and subsequent performance on COMLEX-USA Level 2-PE. Specifically, to investigate what percentage of candidates failed because of misrepresentation on first attempts and how they performed on subsequent administrations, as well as the previous performance patterns of candidates who failed because of misrepresentation on examination retakes. Historical records from NBOME's COMLEX-USA Level 2-PE database (testing cycles 2007-2008, 2008-2009, and 2009-2010) were used to analyze overall and domain pass/fail patterns of candidates who failed at least once because of misrepresentation of clinical findings. Of the 24 candidates who failed because of misrepresentation of postencounter (SOAP) notes, 20 candidates (83%) were first-time examinees. Four candidates (17%) were repeating the examination, 2 of whom were making a third attempt to pass. Among these 20 candidates who failed because of misrepresentation of clinical findings on their first attempt, 19 passed on their next attempt. At the time of study analysis, all but 2 candidates eventually passed the examination in subsequent attempts. Among candidates found to have misrepresented clinical findings on postencounter written documentation on COMLEX-USA Level 2-PE, no pattern existed between their past or subsequent performance with regard to overall or domain pass/fail results. The vast

Modeling the highway transportation of spent fuel

International Nuclear Information System (INIS)

Harrison, I.G.

1986-01-01

There will be a substantial increase in the number of spent fuel shipments on the nation's highway system in the next thirty years. Most of the spent fuel will be moving from reactors to a spent fuel repository. This study develops two models that evaluate the risk and cost of moving the spent fuel. The Minimum Total Transport Risk Model (MTTRM) seeks an efficient solution for this problem by finding the minimum risk path through the network and sending all the spent fuel shipments over this one path. The Equilibrium Transport Risk Model (ETRM) finds an equitable solution by distributing the shipments over a number of paths in the network. This model decreases the risk along individual paths, but increases society's risk because the spent fuel shipments are traveling over more links in the network. The study finds that there is a trade off between path risk and societal risk. As path risk declines, societal risk rises. The cost of shipping also increases as the number of paths expand. The cost and risk of shipping spent fuel from ten reactors to four potential repository sites are evaluated using the MTTRM. The temporary monitored retrievable storage (MRS) facility in Tennessee is found to be the minimum cost and minimum risk solution. When direct shipment to the permanent sites is considered, Deaf Smith, Texas is the least cost and least incident free transport risk location. Yucca Mountain, Nevada is the least risk location when the focus is placed on the potential consequences of an accident

Definitions of fuel poverty: Implications for policy

International Nuclear Information System (INIS)

Moore, Richard

2012-01-01

This paper outlines why the definition of fuel poverty is important in policy formulation and describes how the Government's current definitions evolved from the original concept. It discusses the determination of income and fuel costs and the possibilities for a relative and common European measure. It examines problems inherent in assessing fuel costs as a percentage of income and puts forward the arguments for a ‘budget standard’ approach. The paper illustrates how the size of the problem depends on the definition and chosen threshold and suggests advantages for a rating scale. It illustrates how the income composition and thresholds also govern the distribution of the target populations and the relative importance of the main causal factors, and examines the consequent policy implications. It explores the definition of vulnerable households and the importance of severity and questions whether the UK fuel poverty strategy is targeted at households least able to afford their fuel costs (as the name implies) or primarily those at risk from excess winter and summer mortality and morbidity. Finally, after examining the role of supplementary indicators, it looks at the opportunities for changing the definition and comments on the Government review of the definition and targets. - Highlights: ► There are major failings in the existing official definitions of fuel poverty. ► expressing fuel costs as a percentage of income is a poor indicator of fuel poverty. ► A budget standard approach provides a more consistent, meaningful and fairer measure. ► The scale and nature of the problem changes dramatically with different definitions. ► The definition is crucial to the mix of policies and allocation of resources required.

Proton exchange membrane fuel cells for electrical power generation on-board commercial airplanes

International Nuclear Information System (INIS)

Pratt, Joseph W.; Klebanoff, Leonard E.; Munoz-Ramos, Karina; Akhil, Abbas A.; Curgus, Dita B.; Schenkman, Benjamin L.

2013-01-01

Highlights: ► We examine proton exchange membrane fuel cells on-board commercial airplanes. ► We model the added fuel cell system’s effect on overall airplane performance. ► It is feasible to implement an on-board fuel cell system with current technology. ► Systems that maximize waste heat recovery are the best performing. ► Current PEM and H 2 storage technology results in an airplane performance penalty. -- Abstract: Deployed on a commercial airplane, proton exchange membrane (PEM) fuel cells may offer emissions reductions, thermal efficiency gains, and enable locating the power near the point of use. This work seeks to understand whether on-board fuel cell systems are technically feasible, and, if so, if they could offer a performance advantage for the airplane when using today’s off-the-shelf technology. We also examine the effects of the fuel cell system on airplane performance with (1) different electrical loads, (2) different locations on the airplane, and (3) expected advances in fuel cell and hydrogen storage technologies. Through hardware analysis and thermodynamic simulation, we found that an additional fuel cell system on a commercial airplane is technically feasible using current technology. Although applied to a Boeing 787-type airplane, the method presented is applicable to other airframes as well. Recovery and on-board use of the heat and water that is generated by the fuel cell is an important method to increase the benefit of such a system. The best performance is achieved when the fuel cell is coupled to a load that utilizes the full output of the fuel cell for the entire flight. The effects of location are small and location may be better determined by other considerations such as safety and modularity. Although the PEM fuel cell generates power more efficiently than the gas turbine generators currently used, when considering the effect of the fuel cell system on the airplane’s overall performance we found that an overall

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Log-binomial models: exploring failed convergence.

Science.gov (United States)

Williamson, Tyler; Eliasziw, Misha; Fick, Gordon Hilton

2013-12-13

Relative risk is a summary metric that is commonly used in epidemiological investigations. Increasingly, epidemiologists are using log-binomial models to study the impact of a set of predictor variables on a single binary outcome, as they naturally offer relative risks. However, standard statistical software may report failed convergence when attempting to fit log-binomial models in certain settings. The methods that have been proposed in the literature for dealing with failed convergence use approximate solutions to avoid the issue. This research looks directly at the log-likelihood function for the simplest log-binomial model where failed convergence has been observed, a model with a single linear predictor with three levels. The possible causes of failed convergence are explored and potential solutions are presented for some cases. Among the principal causes is a failure of the fitting algorithm to converge despite the log-likelihood function having a single finite maximum. Despite these limitations, log-binomial models are a viable option for epidemiologists wishing to describe the relationship between a set of predictors and a binary outcome where relative risk is the desired summary measure. Epidemiologists are encouraged to continue to use log-binomial models and advocate for improvements to the fitting algorithms to promote the widespread use of log-binomial models.

Axially alignable nuclear fuel pellets

International Nuclear Information System (INIS)

Johansson, E.B.; Klahn, D.H.; Marlowe, M.O.

1978-01-01

An axially alignable nuclear fuel pellet of the type stacked in end-to-end relationship within a tubular cladding is described. Fuel cladding failures can occur at pellet interface locations due to mechanical interaction between misaligned fuel pellets and the cladding. Mechanical interaction between the cladding and the fuel pellets loads the cladding and causes increased cladding stresses. Nuclear fuel pellets are provided with an end structure that increases plastic deformation of the pellets at the interface between pellets so that lower alignment forces are required to straighten axially misaligned pellets. Plastic deformation of the pellet ends results in less interactions beween the cladding and the fuel pellets and significantly lowers cladding stresses. The geometry of pellets constructed according to the invention also reduces alignment forces required to straighten fuel pellets that are tilted within the cladding. Plastic deformation of the pellets at the pellet interfaces is increased by providing pellets with at least one end face having a centrally-disposed raised area of convex shape so that the mean temperature and shear stress of the contact area is higher than that of prior art pellets

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

Connecticut Transit (CTTRANSIT) Fuel Cell Transit Bus Preliminary Evaluation Results

Science.gov (United States)

2008-10-16

This report describes operations at Connecticut Transit (CTTRANSIT) in Hartford for one prototype fuel cell bus and three new diesel buses operating from the same location. The report discusses the planned fuel cell bus demonstration and equipment us...

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

Real-time locating systems (RTLS) in healthcare: a condensed primer.

Science.gov (United States)

Kamel Boulos, Maged N; Berry, Geoff

2012-06-28

Real-time locating systems (RTLS, also known as real-time location systems) have become an important component of many existing ubiquitous location aware systems. While GPS (global positioning system) has been quite successful as an outdoor real-time locating solution, it fails to repeat this success indoors. A number of RTLS technologies have been used to solve indoor tracking problems. The ability to accurately track the location of assets and individuals indoors has many applications in healthcare. This paper provides a condensed primer of RTLS in healthcare, briefly covering the many options and technologies that are involved, as well as the various possible applications of RTLS in healthcare facilities and their potential benefits, including capital expenditure reduction and workflow and patient throughput improvements. The key to a successful RTLS deployment lies in picking the right RTLS option(s) and solution(s) for the application(s) or problem(s) at hand. Where this application-technology match has not been carefully thought of, any technology will be doomed to failure or to achieving less than optimal results.

Description of gasket failure in a 7 cell PEMFC stack

Energy Technology Data Exchange (ETDEWEB)

Husar, Attila; Serra, Maria [Institut de Robotica i Informatica Industrial, Parc Tecnologic de Barcelona, Edifici U, C. Llorens i Artigas, 4-6, 2a Planta, 08028 Barcelona (Spain); Kunusch, Cristian [Laboratorio de Electronica Industrial Control e Instrumentacion, Facultad de Ingenieria, UNLP (Argentina)

2007-06-10

This article presents the data and the description of a fuel cell stack that failed due to gasket degradation. The fuel cell under study is a 7 cell stack. The unexpected change in several variables such as temperature, pressure and voltage indicated the possible failure of the stack. The stack was monitored over a 6 h period in which data was collected and consequently analyzed to conclude that the fuel cell stack failed due to a crossover leak on the anode inlet port located on the cathode side gasket of cell 2. This stack failure analysis revealed a series of indicators that could be used by a super visional controller in order to initiate a shutdown procedure. (author)

Progress in hydrogen fueled busses

International Nuclear Information System (INIS)

Scott, P.B.; Mazaika, D.M.; Tyler, T.

2004-01-01

'Full text:' The Thor/ISE fuel cell bus has been in demonstration and revenue service during 2002-2003 at sites including SunLine Transit, Chula Vista Transit, Los Angeles County Metropolitan Transit Authority, and AC Transit in Oakland. By taking advantage of ISE's advanced hybrid-electric drive technology, this 30-foot bus operates with a much smaller fuel cell than those used in other buses of this class. Further, stress on the fuel cell is diminished. Based on the exceptional performance of this prototype bus, the transit agencies listed above have concluded that hybrid electric hydrogen fueled buses are attractive. Two types of hydrogen fueled hybrid electric buses will be described: - fuel cell powered, and - HICE (Hydrogen Internal Combustion Engine) This progress report will include: 1. Experience with the Thor/ISE fuel cell bus, including results from revenue service at two transit locations, 2. Design and fabrication status of the advanced fuel cell buses being built for AC Transit and SunLine Transit, 3. Design and fabrication status of the prototype HHICE (Hybrid electric Hydrogen fueled Internal Combustion Engine) bus that uses a Ford hydrogen burning engine, mated to a generator, rather than a fuel cell. Other than the engine, the drive train in the HHICE bus is nearly identical to that of a fuel cell hybrid-electric bus. Canadian participation in the HHICE bus is extensive, it is a New Flyer platform and will be winter tested in Winnipeg. (author)

A model for fission product distribution in CANDU fuel

International Nuclear Information System (INIS)

Muzumdar, A.P.

1983-01-01

This paper describes a model to estimate the distribution of active fission products among the UO 2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I 131 . The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

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.)

Influence of external source location in the reactivity calculation

International Nuclear Information System (INIS)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

2011-01-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

Influence of external source location in the reactivity calculation

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra, E-mail: asilva@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

The logistics and the supply chain in the Juzbado Nuclear Fuel Manufacturing Plant

International Nuclear Information System (INIS)

2005-01-01

The paper describe the logistics and the supply chain in the Juzbado Nuclear Fuel Manufacturing Plant, located in Juzbado in the province of Salamanca. In the the article are described the principal elements in the supply chain and the difficulties of its management derived from the short period for the manufacturing of the nuclear fuel. It's also given a view in relation to the transportation by land sea of the nuclear components, uranium oxide powder and the manufactured fuel. The characteristics of the supply chain are determined by the plant production forecast, by the origin and high technology of the raw materials and by nuclear fuel delivery site locations. (Author)

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

Energy Technology Data Exchange (ETDEWEB)

Lanning, D. D.

1981-09-01

This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527.

A study of UO2 wafer fuel for very high-power research reactors

International Nuclear Information System (INIS)

Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

1983-01-01

The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

A Mobile Robot for Emergency Operation of Fuel Exchange Machine

Energy Technology Data Exchange (ETDEWEB)

Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2007-07-01

A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator.

A Mobile Robot for Emergency Operation of Fuel Exchange Machine

International Nuclear Information System (INIS)

Seo, Yongchil; Lee, Sunguk; Kim, Changhoi; Shin, Hochul; Jung, Seungho; Choi, Changhwan

2007-01-01

A Pressurized Heavy Water Reactor (PHWR) uses a heavy water as the coolant and moderator because it does not attenuate the neutron inside the reactor, which makes it possible to use natural uranium for nuclear fuels. However, since the uranium ratio is too low within the natural uranium, the reactor should be refueled everyday while the reactor is working. For that purpose, there is a fuel exchange machine. However as the time passes by, the durability and reliability become a problem. While the fuel handling machine exchanges the reactor fuel, it can be stuck to the pressure tube attached in the Calandra. Although this kind of situation is rarely happen, it can make the reactor be shutdown for normalizing the operation. Since the refueling is performed while the reactor is working, the radiation level is extremely high and the machine can be located at a high position up to nine meters from the floor, that is, the human worker can not approach the machine, so the fuel handling machine should be released remotely. To cope with this situation, the fuel handling machine has a manual drive mechanism at the rear side of it as shown in the circled images. If the worker can handle these manual drive mechanisms, the fuel handling machine can be released form the pressure tube. The KAERI had developed a long-reach manipulator system with a telescophic mast mechanism which can be deployed in the basement of the reactor room and manipulate the manual lever of the fuel exchange machine. Since the manipulator is located in the basement, there are several problems for its application such that the plug hole should be removed before the operation and the vibration of the mast mechanism make it difficult to locate the end effecter of the manipulator

Selection of a plant location - A case study: Urea production using ...

African Journals Online (AJOL)

Selection of a plant location - A case study: Urea production using Calub Gas. ... The criteria are. amongst others, supply of raw materials and fuel, market proximity, water supply. waste ... To set up · critical factors for Ethiopian scenario

Potential bioethanol feedstock availability around nine locations in the Republic of Ireland

International Nuclear Information System (INIS)

Deverell, R.; McDonnell, K.; Devlin, G.

2009-01-01

The Republic of Ireland, like many other countries is trying to diversify energy sources to counteract environmental, political and social concerns. Bioethanol from domestically grown agricultural crops is an indigenously produced alternative fuel that can potentially go towards meeting the goal of diversified energy supply. The Republic of Ireland's distribution of existing soils and agricultural land-uses limit arable crop land to around 10% of total agricultural area. Demand for land to produce arable crops is expected to decrease, which could open the opportunity for bioethanol production. Bioethanol production plants are required to be of a sufficient scale in order to compete economically with other fuel sources, it is important therefore to determine if enough land exists around potential ethanol plant locations to meet the potential demands for feedstock. This study determines, through the use of a developed GIS based model, the potential quantities of feedstock that is available in the hinterlands of nine locations in the Republic of Ireland. The results indicate that three locations can meet all its feedstock demands using indigenously grown sugarbeet, while only one location can meet its demands using a combination of indigenous wheat and straw as the two locally sourced feedstocks. (author)

The failing firm defence: merger policy and entry

OpenAIRE

Mason, Robin; Weeds, Helen

2003-01-01

This Paper considers the 'failing firm defence'. Under this principle, found in most antitrust jurisdictions, a merger that would otherwise be blocked due to its adverse effect on competition is permitted when the firm to be acquired is a failing firm, and an alternative, less detrimental merger is unavailable. Competition authorities have shown considerable reluctance to accept the failing firm defence, and it has been successfully used in just a handful of cases. The Paper considers the def...

Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

International Nuclear Information System (INIS)

Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

2010-01-01

U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

Coping Styles of Failing Brunei Vocational Students

Science.gov (United States)

Mundia, Lawrence; Salleh, Sallimah

2017-01-01

Purpose: The purpose of this paper is to determine the prevalence of two types of underachieving students (n = 246) (active failing (AF) and passive failing (PF)) in Brunei vocational and technical education (VTE) institutions and their patterns of coping. Design/methodology/approach: The field survey method was used to directly reach many…

Fire, Fuel, and Smoke Program: 2014 Research Accomplishments

Science.gov (United States)

Faith Ann Heinsch; Robin J. Innes; Colin C. Hardy; Kristine M. Lee

2015-01-01

The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support staff in FFS...

Effect of ramp-cavity on hydrogen fueled scramjet combustor

Directory of Open Access Journals (Sweden)

J.V.S. Moorthy

2014-03-01

Full Text Available Sustained combustion and optimization of combustor are the two challenges being faced by combustion scientists working in the area of supersonic combustion. Thorough mixing, lower stagnation pressure losses, positive thrust and sustained combustion are the key issues in the field of supersonic combustion. Special fluid mechanism is required to achieve good mixing. To induce such mechanisms in supersonic inflows, the fuel injectors should be critically shaped incurring less flow losses. Present investigations are focused on the effect of fuel injection scheme on a model scramjet combustor performance. Ramps at supersonic flow generate axial vortices that help in macro-mixing of fuel with air. Interaction of shocks generated by ramps with the fuel stream generates boro-clinic torque at the air & liquid fuel interface, enhancing micro-mixing. Recirculation zones present in cavities increase the residence time of the combustible mixture. Making use of the advantageous features of both, a ramp-cavity combustor is designed. The combustor has two sections. First, constant height section consists of a backward facing step followed by ramps and cavities on both the top and bottom walls. The ramps are located alternately on top and bottom walls. The complete combustor width is utilized for the cavities. The second section of the combustor is diverging area section. This is provided to avoid thermal choking. In the present work gaseous hydrogen is considered as fuel. This study was mainly focused on the mixing characteristics of four different fuel injection locations. It was found that injecting fuel upstream of the ramp was beneficial from fuel spread point of view.

BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

International Nuclear Information System (INIS)

Mildrum, C.M.; Taleyarkhan, R.P.

1987-01-01

In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

Proposed fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Walters, L.C.

1985-01-01

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

International Nuclear Information System (INIS)

Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

2011-01-01

This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

Fuel reprocessing experience in India: Technological and economic considerations

International Nuclear Information System (INIS)

Prasad, A.N.; Kumar, S.V.

1983-01-01

The approach to the reprocessing of irradiated fuel from power reactors in India is conditioned by the non-availability of highly enriched uranium with the consequent need for plutonium for the fast-reactor programme. With this in view, the fuel reprocessing programme in India is developing in stages matching the nuclear power programme. The first plant was set up in Trombay to reprocess the metallic uranium fuel from the research reactor CIRUS. The experience gained in the construction and operation of this plant, and in its subsequent decommissioning and reconstruction, has not only provided the know-how for the design of subsequent plants but has indicated the fruitful areas of research and development for efficient utilization of limited resources. The Trombay plant also handled successfully, on a pilot scale, the reprocessing of irradiated thorium fuel to separate uranium-233. The second plant at Tarapur has been built for reprocessing spent fuels from the power reactors at Tarapur (BWR) and Rajasthan (PHWR). The third plant, at present under design, will reprocess the spent fuels from the power reactors (PHWR) and the Fast Breeder Test Reactor (FBTR) located at Kalpakkam. Through the above approach experience has been acquired which will be useful in the design and construction of even larger plants which will become necessary in the future as the nuclear power programme grows. The strategies considered for the sizing and siting of reprocessing plants extend from the idea of small plants, located at nuclear power station sites, to a large-size central plant, located at an independent site, serving many stations. The paper discusses briefly the experience in reprocessing uranium and thorium fuels and also in decommissioning. An attempt is made to outline the technological and economic aspects which are relevant under different circumstances and which influence the size and siting of the fuel reprocessing plants and the expected lead times for construction

Spent fuel canister docking station

International Nuclear Information System (INIS)

Suikki, M.

2006-01-01

The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

Fission product release as a function of chemistry and fuel morphology

International Nuclear Information System (INIS)

Hobbins, R.R.; Osetek, D.J.; Petti, D.A.; Hagrman, D.L.

1989-01-01

Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are the chemistry within the core and the morphology of the fuel or fuel-bearing debris. This paper reviews the current understanding of the these factors garnered from integral and separate effect experiments and the TMI-2 accident, and provides perspective on the significance of contributing phenomena for the analysis of severe accidents, particularly during the in-vessel phase. Information has been obtained recently on phenomena affecting the release of fission products from fuel and the reactor vessel during the in-vessel melt progression phase of a severe accident. The influence of a number of these phenomena will be reviewed, including fuel chemistry, H 2 /H 2 O ratio, fuel liquefaction, molten pools, and debris beds. 13 refs., 1 fig., 1 tab