WorldWideScience

Sample records for hanaro fuel assembly

  1. Pressure drop measurement for flow-measuring dummy fuel assemblies in HANARO core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heon Il; Chae, Hee Taek; Chung, Heung June [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    In order to characterize the flow distribution of HANARO core, flow-rate measuring dummy fuel assemblies (instrumented dummy fuel assemblies) were to be used in the HANARO commissioning. To do this instrumented dummy fuel assemblies were developed and the calibration tests were conducted in the thermal-hydraulic laboratory. Through this experiment the correlations for 6 instrumented dummy fuel assemblies were derived. The measured total pressure drop for the 36-element dummy fuel assembly was 211 kPa, which meets the design requirement, 209 kPa {+-} 5%. The form loss coefficients for the spacers were re-evaluated and the new correlation was obtained. 7 tabs., 13 figs., 2 refs. (Author).

  2. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  3. Structural Integrity Evaluation of HANARO Fuel Basket

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Kim, Seong Ho; Yoon, Doo Byung

    2005-12-15

    The objective of this report is to perform the seismic analysis and the structural integrity evaluation of HANARO fuel basket and its supporting box beam in water. For these purpose, the in-air and submerged ANSYS finite element models of the HANARO fuel basket and the box beam were developed and the dynamic characteristics were analyzed. The seismic response spectrum analyses of HANARO fuel basket and box beam structure under the design floor response spectrum loads of OBE and SSE were performed. The analysis results show that the stress values of the HANARO fuel basket and the box beam for the seismic loads are within the ASME Code limits. It is also confirmed that the fatigue usage factor is much less than 1.0. Therefore any damage on structural integrity is not expected when the HANARO fuel basket is installed in the upper part of the box beam structure in water.

  4. HANARO fuel irradiation test(I)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Sung; Jeon, Byung Jin; Kim, Hak No; Kang, Byung Wee; Lee, Jong Tak; Kim, Bong Ku; Lim, In Chul; Chae, Hee Taek; Park, Cheol; Lee, Byung Cheol; Oh, Yeon Wo; Ahn, Sang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Various types of test fuel has been reviewed and a 36 element hexagonal bundle design, with only 6 fuel elements at the corners of the peripheral positions and with A1 dummy rods or hollow elements at the remained 30 positions, was selected. The design enables the assembly to get local power boosting and to achieve higher power and faster burnup accumulation than other driver fuel in the core. For the three central positions (CT, IR 1 and IR 2) with maximum neutron flux, three different target burnup levels of {approx} 40, {approx} 65 and 85 {approx} 90 a/o have been chosen for the three test assemblies. One test assembly will be instrumented with thermocouples, neutron and gamma detectors to verify the irradiation condition. The instrumented test assembly has been fabricated in KAERI and other two test assemblies without any instruments were supplied by AECL. Two un-instrumented assemblies was inserted into HANARO at the end of December 1995 and has been under irradiation at IR 1 and IR 2 core positions without any problem. The measured data in the first cycle was conformed well to the predicted results. In the first cycle irradiation, the best estimated maximum power was 100.5 kw/m (MCNP) and the conservative maximum power was 117.7 kw/m. The instrumented test assembly will be irradiated from the second cycle. 5 tabs., 11 figs., 8 refs. (Author).

  5. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  6. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  7. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project ''The Nuclear Fuel Material Development of Research Reactor''. And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,.

  8. Design of the Capsule (13M-01K) for Irradiation of Fuel Cladding Materials in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear-grade zirconium alloys contain more than 95% Zr, and therefore most of their properties are similar to those of pure zirconium. ZIRLO material, used in fuel rod cladding, structural and flow mixing grids, instrumentation tubes, and guide thimbles, increases margin-to-fuel-rod-corrosion limits and enhances fuel assembly structural stability. The demonstrated corrosion resistance and enhanced structural stability of ZIRLO cladding enable longer cycle lengths at higher temperatures without reducing operating margins. An instrumented capsule (13M-01K) was designed and fabricated for evaluation of the neutron irradiation properties of Zirlo material, which is commonly used for cladding of nuclear fuel. This capsule is now being irradiated for 2 cycles at CT test hole of HANARO, which was started at Jan 27 and will be ended at Mar 31, 2014. The structure of the capsule was based on the previous capsule (11M-22K capsule) which was successfully irradiated at the same hole of HANARO. In the capsule, 182 specimens such as tensile specimens of plate type and ring type specimens were placed. Most of them are made of Zirlo, but a few are HANA material that is developed in KAERI. The irradiation test was requested by 4 universities including Dong-Kook and Han-Yang etc. The capsule is composed of 5 layers, each of which had Al holder containing several specimens and an independent electric heater, thermocouples etc. During the irradiation test, temperatures of the specimens and fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule is irradiated for 2 cycles (28 days) at the CT test hole of HANARO of a 30MW thermal output at 350-390 .deg. C up to a fast neutron fluence of 9.6Χ10{sup 20} (n/cm{sup 2}) (E>1.0 Mev). In this capsule, the two kinds of irradiation tests are performed at quite different temperatures. At upper 3 layers of the capsule, specimens for irradiation at high

  9. IRRADIATION DEVICE FOR IRRADIATION TESTING OF COATED PARTICLE FUEL AT HANARO

    Directory of Open Access Journals (Sweden)

    BONG GOO KIM

    2013-12-01

    Full Text Available The Korean Nuclear-Hydrogen Technology Development (NHTD Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about 1.7×1021 n/cm2, PIE (Post-Irradiation Examination of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility. This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

  10. Post irradiation examination of HANARO nucler fuel (KFH-067). Metallographic and density test

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H.; Oho, W. U.; Kim, B. K

    2001-01-01

    The 2nd phase post irradiation examination (PIE) of nuclear fuel for HANARO which is a multi purpose research reactor. Built at KAERI has been performed in order to meet the licensing requirement of it. Changes of microstructure and density between fules with 50 and 80 at% burnup have been examined and compared with each other. The detailed PIE items are as follows; microstructure of diameter relate the swelling of fuel, measurement reaction larger particles and aluminum matrix, measurement of oxide layer thickness of fuel cladding, distribute of U3Si, examination of welding part of end plug examination of end plug part feature, and density measurement of fuel with burnup.

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

    Directory of Open Access Journals (Sweden)

    H.J. RYU

    2013-12-01

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

  12. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  13. Research on water discharge characteristics of PEM fuel cells by using neutron imaging technology at the NRF, HANARO.

    Science.gov (United States)

    Kim, TaeJoo; Sim, CheulMuu; Kim, MooHwan

    2008-05-01

    An investigation into the water discharge characteristics of proton exchange membrane (PEM) fuel cells is carried out by using a feasibility test apparatus and the Neutron Radiography Facility (NRF) at HANARO. The feasibility test apparatus was composed of a distilled water supply line, a compressed air supply line, heating systems, and single PEM fuel cells, which were a 1-parallel serpentine type with a 100 cm(2) active area. Three kinds of methods were used: compressed air supply-only; heating-only; and a combination of the methods of a compressed air supply and heating, respectively. The resultant water discharge characteristics are different according to the applied methods. The compressed air supply only is suitable for removing the water at a flow field and a heating only is suitable for water at the MEA. Therefore, in order to remove all the water at PEM fuel cells, the combination method is needed at the moment.

  14. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

  15. HANARO user support and training

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Shin Ae; Kim, K. Y.; Kim, B. K. (and others)

    2008-06-15

    This project is aimed to support external users for the effective use of HANARO. The total number of projects selected as the beneficiary of the supporting program by MEST was 21 including this project in this year. We supported 2,339 hr measurements for the 31 requests of the 14 projects selected on the field of neutron beam utilization. In the field of materials and nuclear fuel irradiation test the 3 projects were selected and supported for 80 samples. In the fields of neutron activation analysis and radioisotope production the number of selected and supported projects were 1 and 2 respectively. In order to broaden potential user base, maximize instrument utilization, and enhance cooperation with industries, universities and institutes, practice-oriented HANARO user training courses were held for neutron beam utilization and materials and nuclear fuel irradiation fields. In the fields of neutron activation analysis 3 times training courses were held for the university students. The online neutron beam time allocation system was developed and applied successfully for the HRPD in this year. We are planing to apply this system to other neutron beam instruments in the near future. This project is a kind of the user-based supporting program for the maximize of HANARO utilization. The development products and the ideas and suggestions of users obtained through this projects will be collected and applied to the development of next new facilities. Also, by using the 'HANARO utilization and research information management system(HANARO4U)' we construct the research network among users at industries, universities and institutes. This network is expected to increase HANARO utilization and enhance productivity of the facilities.

  16. Promotion of HANARO Utilization for Year 2006

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Kim, H. R.; Jun, B. J. (and others)

    2007-05-15

    To activate the HANARO utilization by expanding HANARO utilization fields, recruiting and training new users. In order to promote the HANARO utilization, the following activities have been performed. The neutron usage fee in HANARO. Achievements of HANARO utilization. Project for Activation of the Research using HANARO. HANARO Symposium. Survey of the HANARO User Satisfaction Index. Operation and Management of HANARO Server. Management of HANARO related committees. Training of HANARO users. Related activities of HANARO publicity. The related activities to activate HANARO utilization have been carried out successfully. This report summarized the detailed activities to activate the HANARO utilization. They will be useful for expanding HANARO utilization in the near future.

  17. Design and fabrication of hafnium tube to control the power of the irradiation test fuel in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H

    2003-05-01

    For the irradiation test at HANARO, non-instrumentation capsule was manufactured and hafnium tube was used to control LHGR of HANARO. Hafnium tube can control the irradiation condition of HANARO similar to that of commercial reactor. Hafnium tube thickness was determined by the LHGR calculated at OR-4 irradiation hole to be installed the non-instrumented capsule. To fabricate the hafnium tube with hafnium plate, the fabrication method was determined by using the hafnium mechanical properties. And the tensile strength of hafnium was confirmed by tensile test. This report is confirmed the LHGR control at the OR-4 and the Hafnium fabrication for in used which the AFPCAP non-instrumented irradiation capsule.

  18. HANARO core channel flow-rate measurement

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heon Il; Chae, Hee Tae; Im, Don Soon; Kim, Seon Duk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    HANARO core consists of 23 hexagonal flow tubes and 16 cylindrical flow tubes. To get the core flow distribution, we used 6 flow-rate measuring dummy fuel assemblies (instrumented dummy fuel assemblies). The differential pressures were measured and converted to flow-rates using the predetermined relationship between AP and flow-rate for each instrumented dummy fuel assemblies. The flow-rate for the cylindrical flow channels shows +-7% relative errors and that for the hexagonal flow channels shows +-3.5% relative errors. Generally the flow-rates of outer core channels show smaller values compared to those of inner core. The channels near to the core inlet pipe and outlet pipes also show somewhat lower flow-rates. For the lower flow channels, the thermal margin was checked by considering complete linear power histories. From the experimental results, the gap flow-rate was estimated to be 49.4 kg/s (cf. design flow of 50 kg/s). 15 tabs., 9 figs., 10 refs. (Author) .new.

  19. Neutron detectors for scattering experiments at HANARO

    Indian Academy of Sciences (India)

    alignment of the monochromator crystal and sample itself. The traditional photo- graphic method is time consuming and does not provide any information on the small variations in intensity. LE PSD is used for beam alignment at all the neutron scattering instruments at HANARO [3]. Figure 3a shows the assembled LE PSD.

  20. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  1. Internal reforming fuel cell assembly with simplified fuel feed

    Science.gov (United States)

    Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

    2001-01-01

    A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

  2. Recent irradiation tests for future nuclear system at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-05-15

    The capsule at HANARO is a device that evaluates the irradiation effects of nuclear materials and fuels, which can reproduce the environment of nuclear power plants and accelerate to reach to the end of life condition. As the integrity assessment and the extension of lifetime of nuclear power plants are recently considered as important issues in Korea, the requirements for irradiation test are gradually being increased. The capacity and capability irradiation tests at HANARO are becoming important because Korea strives to develop SFR (Sodium-cooled Fast Reactor) and VHTR (Very High Temperature Reactor) among the future nuclear system and to export the research reactors and to develop the fusion reactor technology.

  3. Fuel rod assembly to manifold attachment

    Science.gov (United States)

    Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

    1980-01-01

    A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

  4. Simulated nuclear reactor fuel assembly

    Science.gov (United States)

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  5. MOX fuel assembly and reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Motoo; Shimada, Hidemitsu; Kaneto, Kunikazu; Koyama, Jun-ichi; Uchikawa, Sadao [Hitachi Ltd., Tokyo (Japan); Izutsu, Sadayuki; Fujita, Satoshi

    1998-03-10

    MOX fuel assemblies containing fuel rods of mixed oxide (MOX) of uranium and plutonium are loaded to a reactor core of a BWR type reactor. The fuel assembly comprises lattice like arranged fuel rods, one large diameter water rod disposed at the central portion and a channel box surrounding them. An average enrichment degree A of fission plutonium of fuel rods arranged at the outermost layer region and an average enrichment degree B of fission plutonium of fuel rods arranged at the inner layer region satisfy the relation of B/A {>=} 2.2. It is preferable that the average enrichment degree C of fission plutonium of fuel rods arranged at the outermost corner portions and the enrichment degree A satisfy the relation: C/A {<=} 0.5. With such a constitution, even in a case where the MOX fuel assemblies and uranium fuel assemblies are disposed together, thermal margin can be improved. (I.N.)

  6. Fuel Cell Electrodes for Hydrogen-Air Fuel Cell Assemblies.

    Science.gov (United States)

    The report describes the design and evaluation of a hydrogen-air fuel cell module for use in a portable hydrid fuel cell -battery system. The fuel ... cell module consists of a stack of 20 single assemblies. Each assembly contains 2 electrically independent cells with a common electrolyte compartment

  7. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  8. Design, fabrication and irradiation test report on HANARO instrumented capsule (05M-07U) for the researches of universities in 2005

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H.; Choi, M. H.; Cho, M. S.; Son, J. M.; Choi, M. H.; Shin, Y. T.; Park, S. J.

    2006-09-15

    As a part of the 2005 project for an active utilization of HANARO, an instrumented capsule (05M-07U) was designed, fabricated and irradiated for an irradiation test of various unclear materials under irradiation conditions which was requested by external researchers from universities. The basic structure of the 05M-07U capsule was based on the 00M-01U, 01M-05U, 02M-05U, 03M-06U and 04M-07U capsules which had been successfully irradiated in HANARO as part of the 2000, 2001, 2002, 2003 and 2004 projects. However, because of a limited number of specimens and the budget of one university, the remaining space in the capsule was filled with various KAERI specimens for researches on a nuclear core and SMART materials, and parts of a nuclear fuel assembly of KNFC. Various types of specimens such as tensile, Charpy, TEM, hardness, compression and growth specimens made of Zr 702, Ti and Ni alloys, Zirlo, Inconel, STS 316L and Cr-Mo alloys were placed in the capsule. Especially, this capsule was designed to evaluate the nuclear characteristics of the parts of a nuclear fuel assembly and the Ti tubes in HANARO. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 270 ∼ 400 .deg. C up to a fast neutron fluence of 5.7 x 10{sup 20} (n/cm{sup 2}) (E >1.0MeV). The obtained results will be very valuable for the related research of the users.

  9. LMFBR fuel assembly design for HCDA fuel dispersal

    Science.gov (United States)

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  10. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  11. Modular fuel-cell stack assembly

    Science.gov (United States)

    Patel, Pinakin

    2010-07-13

    A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

  12. MOX fuel assembly and reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Hidemitsu; Aoyama, Motoo

    1997-09-25

    A MOX fuel assembly capable of increasing plutonium charge while securing the effect of reactivity control, and a reactor core comprising the same. The assembly comprises fuel rods including MOX fuel rods containing MOX fuel pellets incorporated there and a water rod arranged in the form of a square lattice. The MOX fuel rods include those containing burnable poisons, and the percentage Nfr(%) of the number of the MOX fuel rods containing burnable poisons based on the total number of the fuel rods and the mean weight percentage Cag(%) of the burnable poisons contained in the MOX fuel rods satisfy at least either of the following requirements: -1.7 Cag + 21.8 {<=} Nfr {<=} -4.4 Cag + 56.1, 0.5 {<=} Cag {<=} 5.0. (author) figs.

  13. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  14. Post-Irradiation Examination Test for an Evaluation of In-Core Performance of the Parts of Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. S.; Ryu, W. S.; Choo, Y. S. (and others)

    2007-11-15

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring with fuel cladding, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the guide tube. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 270 {approx} 375 .deg. C up to a fast neutron fluence of 5.7x10{sup 20} n/cm{sup 2}(E>1 MeV). From the spring test of mid grid 1x1 cell and bottom grid plate, the irradiation effects were not found. The irradiation effects on the irradiation growth also were not found. The buckling load of mid grid 1x1 cell tends to increase due to a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. Through these tests of the components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor.

  15. Design requirement on HYPER blanket fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER.

  16. The status and prospects of the irradiation tests at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The integrity assessment of NPP and the extension of lifetime are being considered as important issue, and so the irradiation tests for the structural materials and fuels of NPP should necessarily be conducted. And, the capability of irradiation test is important because Korea participates in developing SFR and VHTR while the development of future nuclear systems is going on to obtain new energy sources all over the world. To effectively support the R and D for new nuclear systems at HANARO, the advanced irradiation technologies such as high-temperature irradiation, instrumentation, and long-term irradiation are being developed. In addition, the researches on the irradiation characteristics of super-conductor materials and new chemical materials are being progressed as a part of fundamental research. Korea is conducting R and D programs relevant to new nuclear systems including research reactor, future nuclear system such as VHTR, SFR and fusion reactor system. In addition, research on the irradiation characteristics of super-conductor materials and new electronic materials is being conducted as a part of fundamental research. Irradiation tests in HANARO are mostly related to the R and D relevant to the ageing management and safety evaluation of NPP and development of the future nuclear system and production of design data of research reactor. The HANARO irradiation capsule system has been developed and actively utilized for the irradiation testing of fuels and materials. The irradiation tests of materials up to the present have been performed usually at temperatures below 300 .deg. C, at which the RPV of PWR materials is being operated. As the irradiation tests of the materials and fuels for a Gen-IV nuclear system are recently required and the development of new research reactors is going on, the capsules for high- and low-temperature irradiation are being developed at HANARO.

  17. Polymer electrolyte membrane assembly for fuel cells

    Science.gov (United States)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  18. In-Service Inspection for HANARO Reactor Structure

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Sang Ik; Lee, Young Sub; Lee, Jung Hee; Jung, Hoan Sung; Cho, Yeong Garp [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    As results of the visual inspection (VT-3) on HANARO, there was no defect such as a crack, corrosion, abrasion, erosion, and physical or chemical deformation on the welded parts and other surfaces of the reactor structures and its parts. Also, the bolts and steel wire, which was tied as an indicator whether got released or not, are well being kept its integrity. Throughout this visual inspection results, it turned out that HANARO is well maintaining its integrity and initial condition after installation and assembly in 1995. This inspection was done as a second change in Sep. 2006 according to the long term inspection plan for the safety class component by ASME Sec. XI article IWB and HANARO procedure. The next five year inspection shall be done no later than the end of 2011. This inspection results such as photos taken inside the reactor pool will be helpful to find out a solution for some problem, which may cause a view point of the reactor structure and its parts during operating

  19. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  20. MOX fuel assembly and reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

    1998-06-26

    The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

  1. Low-Temperature and Long-Term Irradiation Testing of HANARO Instrumented Capsule

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoontaek; Park, Sengjae; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this paper, the low-temperature and long-term irradiation capsule technology developed for the irradiation testing of research reactor materials at HANARO are described. A new capsule and capsule system for long-term irradiation at low temperature was designed, fabricated, and irradiated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of a research reactor for the National Project of 'Research Reactor Development'. Two capsules were first designed and fabricated to irradiate materials at low temperature (36-56 .deg .C) for a long cycle of 8 irradiation cycles at HANARO. The safety of the new irradiation capsule and capsule system was thoroughly evaluated through the out-pile and in-pile testing. The new capsule was successfully irradiated for up to 8 cycles at HANARO. HANARO instrumented irradiation capsules have been actively used for the irradiation of nuclear fuels and materials. Capsule technology was basically developed for irradiation testing under a commercial reactor operation environment. Most irradiation testing using capsules has been performed at specimen temperatures of 250-500 .deg. C within 4 reactor operation cycles (about 100 days) at HANARO. Recently, as a part of the research reactor development's project, irradiation testing of materials used as reflector materials in a research reactor such as graphite, beryllium, and zircaloy-4 was required up to 8 reactor operation cycles at low temperature (<100 .deg. C) of specimens.

  2. Membrane electrode assembly for a fuel cell

    Science.gov (United States)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  3. Measurement Protocols for Optimized Fuel Assembly Tags

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-11-01

    This report describes the measurement protocols for optimized tags that can be applied to standard fuel assemblies used in light water reactors. This report describes work performed by the authors at Pacific Northwest National Laboratory for NA-22 as part of research to identify specific signatures that can be developed to support counter-proliferation technologies.

  4. Reconstitution of fuel assemblies and core components

    Energy Technology Data Exchange (ETDEWEB)

    Langenberge, J.

    2012-07-01

    Due to AREVAS experience and big portfolio of techniques, reconstitution of fuel assemblies and core components at light water reactors is possible within a reasonable time frame and with interesting cost benefit Customer feedback indicates the sustainability of such reconstitutions. As a result, a long-term maintenance of value can be assured and early waste disposal can be avoided.

  5. CONTRIBUTION OF HANARO IRRADIATION TECHNOLOGIES TO NATIONAL NUCLEAR R&D

    Directory of Open Access Journals (Sweden)

    KEE NAM CHOO

    2014-08-01

    Full Text Available HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI. Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL facilities, capsule irradiation facilities, and neutron transmutation doping (NTD facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART, research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.

  6. Contribution of HANARO irradiation technologies to national nuclear R and D

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Yang, Sung Woo; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-08-15

    HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R and D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R and D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.

  7. Utilization of the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kyung Bae; Yoon, Byung Mok; Chung, Yong Sam; Han, Hyun Soo; Bang, Hong Sik; Shin, Byung Chul; Cho, Woon Gap; Kim, Young Mi; Park, Ul Jae; Moon, Jong Hwa; Kim, Joo Young; Park, Choon Deuk; Seo, Choon Ha; Lee, Young Koo; Shin, Hun Young; Han, Kwang Hee; Chung, Young Joo; Park, Woong Woo; Kim, Dong Soo; Jun, Sang Soo; Hong, Soon Buk; Min, Eun Sook; Jin, Joon Ha; Kim, Ki Yup; Nho, YOung Chang; Lee, Young Keun; Jung, Sung Hee; Park, Soon Cheol; Na, Bong Joo; Kim, Jae Ho; Paik, Sam Tae; Park, Sang Jun; Suh, Doo Whan; Lee, Jae Poong; Oh, Sam Yong; Youm, Kee Hong; Kim, Jong Kook; Gu, Jung Min; Sin, Ho Chul

    1997-01-01

    Radioisotope facilities were operated to support research, development and production of radioisotopes. Irradiation facilities were operated to offer irradiation service to the clients. Even though intensive periodical checks and some essential maintenance had been carried out the results of RI facility maintenance seemed to be not much satisfactory due mainly to the shortage of skilled manpower and maintenance expenses. High intensity {gamma} irradiator was normally operated 165 times for overall 2,670 hrs, and the low intensity {gamma} irradiator was operated 89 times for overall 367 hrs, respectively. The reactor shutdown report of TRIGA Mark-II and Mark-III were submitted to MOST and the spent fuels from the core were transferred to storage racks in reactor pool for decommissioning which will be started in 1997. (author). 5 tabs., 3 figs.

  8. FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.

    1958-04-29

    A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

  9. Development of the Homepage for the HANARO information platform

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Park, K. B.; Kim, H. R. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    I - Object and Importance- Following in the footsteps of internationalization and the information-oriented society, we need to open the HANARO to the public, and to serve more detailed, accurate, and varied information rapidly via the internet to enhance the HANARO Utilization efficiency. II- Scope and Contents- We develope the HANARO Homepage using ORACLE DBMS and APACJE, it has function as an information platform for HANARO. In this report, we describe its design, implementation, management and operation. III-Result -The HANARO homepage was developed successfully, and it was opened to HANARO users from March 20, 2002 through the KAERI-Net. IV- Proposal for Application-The data of HANARO should be processed and organized systematically for its better utilization. A supplementation of the functions is needed and the HANARO web should be operated practically with maximum efficiency and advertise the activities locally and internationally. 3 refs., 11 figs., 9 tabs. (Author)

  10. Selected Isotopes for Optimized Fuel Assembly Tags

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-10-01

    In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

  11. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  12. Validity and Utilization of the Out-Pile Testing Facilities at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee-Nam; Cho, Man-Soon; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jun, Byung-Hyuk; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Various neutron irradiation facilities such as rabbit irradiation facilities, loop facilities and the capsule irradiation facilities for irradiation tests of nuclear materials, fuels and radioisotope products have been developed at HANARO. Among these irradiation facilities, the capsule is the most useful device for coping with the various test requirements at HANARO. To support the national research and development programs on nuclear reactors and the nuclear fuel cycle technology in Korea, new irradiation capsules have been developed and actively utilized for the irradiation tests requested by numerous users. The environmental conditions for these reactors are generally beyond present day reactor technology, especially regarding the higher neutron fluence and higher operating temperature. To effectively support the national R and Ds relevant to the future nuclear systems, the development of advanced irradiation technologies concerning higher neutron fluence and irradiation temperature are being preferentially developed at HANARO. The utilization of the out-pile testing facilities to satisfy the criteria of safety evaluation for a new device installed in the core of HANARO was summarized. In addition, the validity of the out-pile testing facilities was evaluated and proved to be effective for verifying the integrity of irradiation capsule.

  13. Development of alternative fuel assembly for WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Solonin, M.I.; Bibilashvili, Y.K.; Sokolov, N.B. [Vserossijskij Nauchno-Issledovatel`skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Panyushkin, A.K.; Tsibulia, V. [Mashinostroitelniy Zavod, Karl Marx st. 12, Electrostal 144001 (Russian Federation); Samoylov, O.B.; Kurilev, V.B.; Kuul, V.S.; Kaidalov, V.B.; Peskov, R.A.; Ershov, V.F. [OKBM, N. Novgorod 603074 (Russian Federation)

    1997-10-01

    An alternative design of fuel assembly has been developed for the WWER-1000 reactor with the aim of assuring a geometrical stability of the core during operation. The fuel assembly provides enhanced safety and substantial improvement in the WWER-1000 fuel cycle economics. (orig.)

  14. Coupled analysis of the HPLWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Monti, Lanfranco; Schulenberg, Thomas; Starflinger, Joerg [Forschungszentrum Karlsruhe (Germany). Inst. for Nuclear and Energy Technologies

    2008-07-01

    The High Performance Light Water Reactor (HPLWR) [1] is an innovative type of reactor cooled and moderated with water at super-critical pressure. Due to the nominal operation pressure of 25 MPa no two-phase transition occurs and thus the issues related to the nucleate boiling are eliminated permitting a larger coolant temperature rise than for common LWRs and hence the thermodynamic efficiency of the power plant is also increased. The target water heat up, from 550K to 770K, is split into three parts and, because of the associated strong density reduction, the presence of additional in-core flow paths for high density water is required in order to achieve a thermal neutron spectrum. Since the water densities determine the neutron spectrum and hence the power generation which in turn may change the water temperatures and densities, a coupled Neutronic / Thermal-Hydraulic analysis is mandatory even for steady-state conditions as already shown in [2] and [3]. Two stand alone codes where chosen and coupled trough external procedures which exchange the coupling parameters, namely the water density together with the fuel temperature from the thermalhydraulic analysis and the axial power distribution from the neutronic one. This paper will present preliminary results obtained for a steady state coupled analysis of one single HPLWR fuel assembly together with the computational methods developed. (orig.)

  15. Combustor with two stage primary fuel assembly

    Science.gov (United States)

    Sharifi, Mehran; Zolyomi, Wendel; Whidden, Graydon Lane

    2000-01-01

    A combustor for a gas turbine having first and second passages for pre-mixing primary fuel and air supplied to a primary combustion zone. The flow of fuel to the first and second pre-mixing passages is separately regulated using a single annular fuel distribution ring having first and second row of fuel discharge ports. The interior portion of the fuel distribution ring is divided by a baffle into first and second fuel distribution manifolds and is located upstream of the inlets to the two pre-mixing passages. The annular fuel distribution ring is supplied with fuel by an annular fuel supply manifold, the interior portion of which is divided by a baffle into first and second fuel supply manifolds. A first flow of fuel is regulated by a first control valve and directed to the first fuel supply manifold, from which the fuel is distributed to first fuel supply tubes that direct it to the first fuel distribution manifold. From the first fuel distribution manifold, the first flow of fuel is distributed to the first row of fuel discharge ports, which direct it into the first pre-mixing passage. A second flow of fuel is regulated by a second control valve and directed to the second fuel supply manifold, from which the fuel is distributed to second fuel supply tubes that direct it to the second fuel distribution manifold. From the second fuel distribution manifold, the second flow of fuel is distributed to the second row of fuel discharge ports, which direct it into the second pre-mixing passage.

  16. Fuel injection assembly for use in turbine engines and method of assembling same

    Science.gov (United States)

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-03-24

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

  17. Fuel burner and combustor assembly for a gas turbine engine

    Science.gov (United States)

    Leto, Anthony

    1983-01-01

    A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

  18. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

  19. Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

    Science.gov (United States)

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

    2017-09-01

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

  20. Fabrication and characterisation of hydrogen fuel cell membrane electrode assemblies

    CSIR Research Space (South Africa)

    Mathe

    2006-09-01

    Full Text Available Proton-exchange membrane fuel cells (PEMFC) have been receiving attention owing to their highly attractive properties as a power source for both stationary and mobile applications. Fabrication of Membrane Electrode Assemblies (MEAs) is a key...

  1. Transport of fresh MOX fuel assemblies for MONJU initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kurakami, Jun-ichi [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Matsushima, Hideya

    1994-12-01

    Transport of fresh MOX fuel assemblies for prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As much as 205 fresh MOX fuel assemblies were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for type B(U) packaging. Moreover, this packaging design features such advanced technologies as high-performance neutron shielding material and automatic hold-down mechanism for fuel assemblies. Every effort was paid to execute safe transport in conjunction with the cooperation of every competent organization. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (author).

  2. Fuel assembly assessment from CVD image analysis: A feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Lindsay, C.S.; Lindblad, T. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Physics

    1997-05-01

    The Swedish Nuclear Inspectorate commissioned a feasibility study of automatic assessment of fuel assemblies from images obtained with the digital Cerenkov viewing device currently in development. The goal is to assist the IAEA inspectors in evaluating the fuel since they typically have only a few seconds to inspect an assembly. We report results here in two main areas: Investigation of basic image processing and recognition techniques needed to enhance the images and find the assembly in the image; Study of the properties of the distributions of light from the assemblies to determine whether they provide unique signatures for different burn-up and cooling times for real fuel or indicate presence of non-fuel. 8 refs, 27 figs.

  3. The Model of Temperature Dynamics of Pulsed Fuel Assembly

    CERN Document Server

    Bondarchenko, E A; Popov, A K

    2002-01-01

    Heat exchange process differential equations are considered for a subcritical fuel assembly with an injector. The equations are obtained by means of the use of the Hermit polynomial. The model is created for modelling of temperature transitional processes. The parameters and dynamics are estimated for hypothetical fuel assembly consisting of real mountings: the powerful proton accelerator and the reactor IBR-2 core at its subcritica l state.

  4. Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

    2014-02-01

    This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

  5. Guideline for design requirement on KALIMER driver fuel assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.; Lim, J. S.; Ryu, W. S.; Min, B. T.; Kim, D. H.; Kim, Y. J.

    1998-01-01

    This document describes design requirements which are needs for designing the driver fuel assembly duct of the KALIMER as design guidance. The driver fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way, fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle attaches to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the coolant inlet. It contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The design requirements are intended to be used for the design of the driver fuel assembly duct of the KALIMER. (author). 16 refs., 4 figs.

  6. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  7. Nuclear Fuel Assembly Assessment Project and Image Categorization

    Energy Technology Data Exchange (ETDEWEB)

    Lindsey, C.S.; Lindblad, T.; Waldemark, K. [Royal Inst. of Tech., Stockholm (Sweden); Hildingsson, Lars [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    1998-07-01

    A project has been underway to add digital imaging and processing to the inspection of nuclear fuel by the International Atomic Energy Agency. The ultimate goals are to provide the inspector not only with the advantages of Ccd imaging, such as high sensitivity and digital image enhancements, but also with an intelligent agent that can analyze the images and provide useful information about the fuel assemblies in real time. The project is still in the early stages and several interesting sub-projects have been inspired. Here we give first a review of the work on the fuel assembly image analysis and then give a brief status report on one of these sub-projects that concerns automatic categorization of fuel assembly images. The technique could be of benefit to the general challenge of image categorization

  8. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

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

    2016-10-01

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

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

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

  10. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  11. Fuel assembly design study for a reactor with supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Hofmeister, J. [RWE Power AG, Huyssenallee 2, D-45128 Essen (Germany); Waata, C. [ANSYS Germany GmbH, Staudenfeldweg 12, D-83624 Otterfing (Germany); Starflinger, J. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany); Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany)]. E-mail: thomas.schulenberg@iket.fzk.de; Laurien, E. [University of Stuttgart, Institute for Nuclear Technology and Energy Systems (IKE), Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2007-08-15

    The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor.

  12. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  13. Development of cancer therapy facility of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Byung Jin; Hwang, S. Y.; Kim, M. J. and others

    2000-04-01

    Facilities of the research and clinical treatments of neutron capture therapy using HANARO are developed, and they are ready to install. They are BNCT irradiation facility and prompt gamma neutron activatiion analysis facility. Since every horizontal neutron facility of HANARO is long and narrow tangential beam tube, it is analysed that sufficient epithermal neutrons for the BNCT cannot be obtained but sufficient thermal neutrons can be obtained by a filter composed of silicon and bismuth single crystals. Since the thermal neutron penetaration increases significantly when the crystals are cooled, a filter cooled by liquid nitrogen is developed. So as to avoid interference with the reactor operation, a water shutter is developed. The irradiation room is designed for the temporary surgical operation as well. Handling tools to remove activated beam port plug and to install water shutter and filter are developed. The basic structure of the irradiation room is already installed and most of other parts are ready to install. Since no free beam port is available for the prompt gamma neutron activation analysis, a method obtaining almost pure thermal neutrons by the vertical diffraction of extra beam for the polarized neutron spectrometer is developed. This method is confirmed by analysis and experiments to give high enough neutron beam. Equipment and devices are provided to install this facility.

  14. Storage, transportation and disposal system for used nuclear fuel assemblies

    Science.gov (United States)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  15. Design requirement on KALIMER blanket fuel assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs.

  16. Nuclear imaging of the fuel assembly in ignition experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Seguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  17. Fuel assembly gripping device using self-locking mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Lee, G. M.; Choi, S.; Kim, K. S.; Kim, T. W.; Jeong, K. H.; Park, K. B.; Chang, M. H

    1999-07-01

    This report presents an actuating principles and structure for two kind of the fuel assembly gripping devices (Gripper-A, B) developed for SMART. The main components of these grippers are push bundle, rotation bundle, upper guide tube and chuck assembly. The rope attached to winch system on moving cask hangs gripper's push bundle. Due to a down-and-up operation of winch system, the push bundle pushes crown teeth shaped rotation bundle and then it is pushed down and rotated counter clockwise. The push-and-pull sequential operation of push bundle makes the rotation bundle is pushed, rotated and returned, moreover it makes the chuck assembly is expanded or shrunk. The expansion and shrinkage motion of chuck assembly makes the gripper latch and release the fuel assembly. Gripper-A suits for the handling of the fuel assembly with square shaped latching hole. Otherwise Gripper-B suits for a circular shaped latching hole. (author). 5 refs., 20 figs.

  18. Research for the concept of Hanaro cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Oong; Cho, M. S.; Lee, M. W.; Sohn, J. M.; Park, K. N.; Park, S. H.; Yang, S. Y.; Kang, S. H.; Yang, S. H.; Chang, J. H.; Lee, Y. W.; Chang, C. I.; Cho, Y. S.

    1997-09-01

    This report consists of two parts, one is the conceptual design performed on the collaboration work with PNPI Russia and another is review of Hanaro CNS conceptual design report by Technicatome France, both of which are contained at vol. I and vol. II. representatively. In the vol. I, the analysis for the status of technology development, the technical characteristics of CNS is included, and the conceptual design of Hanaro cold neutron source is contained to establish the concept suitable to Hanaro. The cold neutron experimental facilities, first of all, have been selected to propose the future direction of physics concerning properties of the matter at Korea. And neutron guide tubes, the experimental hall and cold neutron source appropriate to these devices have been selected and design has been reviewed in view of securing safety and installing at Hanaro. (author). 38 refs., 49 tabs., 17 figs.

  19. Development of Reactivity Calculation Code for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. G.; Kim, M. S. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Reactivity of the reactor core is measured by a multi-channel wide range reactivity computer (or called reactivity meter), which uses current signals from the compensated ion chamber (CIC) mounted on the courtside wall of the reflector tank in the pool [1]. Because there were a few difficulties in operating the reactivity meter in the MS-DOS environment, some researches have been carried out to improve and upgrade it on the Windows environment [2]. Nevertheless, it is still hard for reactor operators to immediately check the time-dependent reactivity in case of power excursion because of some limitations such as aging of devices and compatibility issues. In this study, a simple off-line tool which can estimate the timedependent reactivity by using the fission chamber signals has been developed, and utilized to the case of loose parts of dummy rod in the 86-2th cycles of HANARO. In order to check the reactivity quickly for reactor operators, the inverse kinetic quations have been incorporated, and several useful functions have been implemented to the code. in the case of 86-2th cycles of HANARO, the developed code showed good performance to estimate time dependent reactivity. In the future, the on line analysis modules will be implanted to the code with upgrade of the measrement equipment such as current meters and data acquisition devices. Additionally, reactivity will be estimated by using the reactivity meter in the MS-DOS enviroment, and the new Windows version, for the verification of the developed code.

  20. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

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

  2. Conceptual design of ASTRID fuel sub-assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Beck, Thierry, E-mail: thierry.beck@cea.fr [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Blanc, Victor; Escleine, Jean-Michel [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Haubensack, David [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France); Pelletier, Michel; Phelip, Mayeul [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Perrin, Benoît [AREVA-NP, 10 rue J. Récamier, 69456 Lyon Cedex 06 (France); Venard, Christophe [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France)

    2017-04-15

    Highlights: • The fuel sub-assembly design for the ASTRID CFV core is described. • Innovative design choices have been made to comply with the GEN IV objectives. • The heterogeneous and the large fuel pins contribute to a low sodium void worth. • The upper neutron shielding is removable from the S/A head before washing. - Abstract: The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project has reached the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors. The main objective to improve safety compared with current GEN II or III reactors led to a core design that demonstrates intrinsically safe behaviour. A negative sodium void worth is achieved thanks to a new fuel sub-assembly design including (U,Pu)O{sub 2} and UO{sub 2} axially heterogeneous fuel pins, a large cladding/small spacer wire bundle, a sodium plenum above the fuel pins, and upper neutron shielding with both enriched and natural boron carbide (B{sub 4}C) which also maintain a low secondary sodium activity level. As these Na-bonded B{sub 4}C pins can lead to the retention of unacceptable amounts of sodium, the whole upper neutron shielding has been made removable on-line through the sub-assembly head just before the washing operations. Finite elements calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core flowering and compaction events. More generally, all design choices for ASTRID have been made with the permanent objective of minimising the sub-assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel sub-assembly design for

  3. An Enhancement of Visual Test Performance for Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Jung Cheol [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2009-05-15

    In the overhaul period of the nuclear power plant, integrity of the neutron-irradiated fuel assembly is evaluated. Nuclear regulations require that nuclear power plants meet the design, operation, and inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B and PV). Section XI of the ASME B and PV Code provides the specific requirements for inspecting the systems, structures, and components; Section V of the ASME Code provides requirements for inspection methods, including volumetric (e.g., ultrasonic testing), surface (e.g., eddy current testing), and visual testing (VT). Visual testing of neutron irradiated fuel assembly is conducted generally for a variety of purposes, for example to detect discontinuities and imperfections on the surface of fuel rods, to detect evidence of leakage from end-cap welds, and to determine the general mechanical and structural condition of one. VT is performed remotely using video camera. As the neutron-irradiated fuel assembly is a high dose-rate gamma-ray source, approximately a few kGy, radiation hardened underwater camera is used in the VT of the fuel assembly. Utilities today follow the EPRI guidelines for VT-1 tests on nuclear components (BWR Vessel and Internals Project-3 1995). The VT-1 guidelines specify which areas around a weld should be examined, how to measure the sizes of indications found, and how to test the resolving power of the visual equipment used for the test. The EPRI guidelines use two 12{mu}m (0.0005-in.) wires or notches as a resolution calibration standard. According to the EPRI guidelines (BWRVIP-03 1995), the camera systems employed were marginally able to detect the 0.0005-inch (12-{mu}m) diameter wire on a steel background. In the some future, it is required that the VT of nuclear fuel assembly follows the EPRI VT-1 guideline. In order to meet the VT-1 guideline, any system used in VT (ranging from the naked eye to a digital closed-circuit TV

  4. Fuel assembly design for APR1400 with low CBC

    Energy Technology Data Exchange (ETDEWEB)

    Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  5. Fuel assembly design for APR1400 with low CBC

    Science.gov (United States)

    Hah, Chang Joo

    2015-04-01

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to ΔkTARGET. A set of new designed fuel assembly satisfies an objective function, min [f =∑i (ΔkF A-Δki ) ] , and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to ΔkTARGET as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  6. Control assembly for controlling a fuel cell system during shutdown and restart

    Science.gov (United States)

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  7. BNCT Technology Development on HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ki Jung; Park, Kyung Bae; Whang, Seung Ryul; Kim, Myong Seop

    2007-06-15

    So as to establish the biological effects of BNCT in the HANARO Reactor, biological damages in cells and animals with treatment of boron/neutron were investigated. And 124I-BPA animal PET image, analysis technology of the boron contents in the mouse tissues by ICP-AES was established. A Standard clinical protocol, a toxicity evaluation report and an efficacy investigation report of BNCT has been developed. Based on these data, the primary permission of clinical application was acquired through IRB of our hospital. Three cases of pre-clinical experiment for boron distribution and two cases of medium-sized animal simulation experiment using cat with verifying for 2 months after BNCT was performed and so the clinical demonstration with a patient was prepared. Also neutron flux, fast neutron flux and gamma ray dose of BNCT facility were calculated and these data will be utilized good informations for clinical trials and further BNCT research. For the new synthesis of a boron compound, o-carboranyl ethylamine, o-carboranylenepiperidine, o-carboranyl-THIQ and o-carboranyl-s-triazine derivatives were synthesized. Among them, boron uptake in the cancer cell of the triazine derivative was about 25 times than that of BPA and so these three synthesized methods of new boron compounds were patented.

  8. Development of a thermal-hydraulic analysis code for annular fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vishnoi, A.K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Reseach Centre (BARC), Mumbai, Maharashtra (India)

    2012-03-15

    In this work a detailed study of the annular fuel has been carried out. A thermal hydraulics code, ANUFAN (Annular Fuel Analysis), based on the bundle average method, capable of modeling both internally and externally cooled annular fuel pins is developed. Code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. Heat transfer fraction difference between ANUFAN and RELAP was found about 1.7%. Analysis of a 54 - fuel rod assembly is carried out with 36 and 45 numbers of annular fuel pins keeping the same channel size and bundle power as of the solid fuel assembly. Fuel pin maximum temperature of the annular fuel is found much less than the solid fuel. MCHFR value for annular fuel is found much higher compared to that of the solid fuel of 54 - fuel rod assembly. The full paper covers the details of the computer code, the analysis carried out and the results obtained. (orig.)

  9. Hydrogen-Oxygen Reaction Assessment in the HANARO Cold Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jung Woon; Kim, Hark Rho; Lee, Kye Hong; Han, Young Soo; Kim, Young Ki; Kim, Seok Hoon; Jeong, Jong Tae

    2006-04-15

    Liquid hydrogen, filled in the moderator cell of the in-pool assembly (IPA), is selected as a moderator to moderate thermal neutrons into cold neutrons for the HANARO Cold Neutron Source. Since the IPA will be installed in the vertical CN hole of the reflector tank at HANARO, the vacuum chamber (VC), the pressure boundary against the reactor, should withstand the detonation pressure so as to avoid any physical damage on the reactor under the hydrogen-oxygen chemical reaction. Accordingly, not only will the vacuum chamber be designed to keep its integrity against the hydrogen accident, but also the hydrogen and vacuum system will be designed with the leak-tight concept and also designed to be surrounded by the inert gas blanket system to prevent any air intrusion into the system. Also, in order to confirm the design concept of the CNS as well as VC integrity against the hydrogen accident, the hydrogen-oxygen chemical reaction is evaluated in this report by several methodologies: AICC methodology, Equivalent TNT detonation methodology, Explosion test result, and Calculation of VC strain under the maximum reflected explosion load.

  10. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  11. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  12. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  13. A spacer grid hysteretic model for the structural analysis of spent fuel assemblies under impact

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, P.R.; Kurkchubasche, I. [ANATECH Research Corp., San Diego, CA (United States); Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States)

    1991-12-31

    This paper presents a methodology for determining the response of spent fuel assembly spacer grids subjected to transport cask impact loading. The spacer grids and their interaction with rod-to-rod loading are the most critical components governing the structural response of spent fuel assemblies. The purpose of calculating the assembly response is to determine the resistance to failure of spent fuel during regulatory transport. The failure frequency computed from these analyses is used in calculating category B spent fuel cask containment source term leakage rates for licensing calculations. Without defensible fuel rod failure frequency prediction calculations, assumptions of 100% fuel failure must be made, leading to leak tight cask design requirements.

  14. Intervention of malignant biliary obstruction with Hanaro spiral stent

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sung Wook; Choo, Sung Wook; Pyeun, Yong Seon and others [Sungkyunkwan Univ. School of Medicine, Seoul (Korea, Republic of)

    1999-08-01

    To evaluate the long-term patency of the Hanaro spiral stent (Solco Intermed, Seoul, Korea) when used as a palliative in patients with inoperable malignant biliary obstruction. Between April 1996 and July 1998, 39 patients with malignant biliary obstruction underwent percutaneous placement of 48 Hanaro spiral stents. The causes of obstruction were bile duct carcinoma (n=18), pancreatic carcinoma (n=8), metastatic lymphadenopathy (n=5), gallbladder carcinoma (n=5), hepatocellular carcinoma (n=1) and other tumors (n=2). Using the kaplan-Meier method, patient survival and stet patency rates were estimated with regard to level of obstruction. As regards stent insertion, there was no technical failure. Overall 25- and 50-week survival rates for the entire patient group were 50 % and 11 %, respectively, while overall stent patency rates at 25 and 50 weeks were 42 % and 11 %, respectively. Twenty-five-week stent patency rates in patients with common bile duct (CBD) and hilar obstruction were 51 % and 18 %, respectively. The stent patency rates in the CBD obstruction group was significantly higher than that in the hilar obstruction group (p<0.05). In patients with CBD obstruction, the clinical efficacy of Hanaro spiral stent was superior to that in patients with hilar obstruction. However, Hanaro spiral stents showed a lower patency rate with regard to patient survival, and further investigation is required.

  15. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  16. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  17. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  18. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  19. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  20. Improvement of the Display Design for Upgraded HANARO Operator Workstations

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sang Hoon; Jung, H. S.; Kim, Y. K.; Choi, Y. S.; Choi, H. Y.; Kim, S. J.; In, W. H.

    2007-12-15

    While the existed HANARO Operator-Workstation(OWS) system was upgraded, the display design of OWS was modified altogether and improved more familiarly to operators according to the newly applied style-guide. It was verified that this upgraded system was reliably compatible with the existed Human-Machine Interface(HMI) surroundings and the improved points comparing before were described as newly designed OWS displays were divided by each functions

  1. Basic Design of the Cold Neutron Research Facility in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hark Rho; Lee, K. H.; Kim, Y. K. (and others)

    2005-09-15

    The HANARO Cold Neutron Research Facility (CNRF) Project has been embarked in July 2003. The CNRF project has selected as one of the radiation technology development project by National Science and Technology Committee in June 2002. In this report, the output of the second project year is summarized as a basic design of cold neutron source and related systems, neutron guide, and neutron scattering instruments.

  2. A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Cho, M. S.; Choo, K. N.; Shin, Y. T.; Sohn, J. M.; Park, S. J.; Kim, B. G

    2008-01-15

    1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

  3. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  4. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Goodsell, Alison Victoria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  5. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  6. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    Science.gov (United States)

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  7. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  8. NUCLEAR DATA UNCERTAINTY AND SENSITIVITY ANALYSIS WITH XSUSA FOR FUEL ASSEMBLY DEPLETION CALCULATIONS

    National Research Council Canada - National Science Library

    Zwermann, W; Aures, A; Gallner, L; Hannstein, V; Krzykacz-Hausmann, B; Velkov, K; Martinez, J.S

    2014-01-01

    Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II...

  9. Development of numerical models for Monte Carlo simulations of Th-Pb fuel assembly

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2017-01-01

    Full Text Available The thorium-uranium fuel cycle is a promising alternative against uranium-plutonium fuel cycle, but it demands many advanced research before starting its industrial application in commercial nuclear reactors. The paper presents the development of the thorium-lead (Th-Pb fuel assembly numerical models for the integral irradiation experiments. The Th-Pb assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design of the assembly allows different combinations of rods for various types of irradiations and experimental measurements. The numerical model of the Th-Pb assembly was designed for the numerical simulations with the continuous energy Monte Carlo Burnup code (MCB implemented on the supercomputer Prometheus of the Academic Computer Centre Cyfronet AGH.

  10. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  11. Efficiency measurement of spin flippers developed at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Doe, Changwoo K.; Jang, Hyung Sik; Choi, Sung Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Moon, Myoung Kook; Lee, Chang Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    Key components of neutron spin echo instrument have been developed at ST-1 beam port of the HANARO reactor. Polarized thermal neutrons produced by (200) reflection plane of Co0.92Fe0.08 ferromagnetic single crystal were used. The spin flipper efficiency and the polarizer-analyzer efficiency product were measured using two pi spin flipper method. Experiment results showed high enough spin flipper efficiency (f{approx}0.98) while the polarizer-analyzer efficiency product was estimated rather low.

  12. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-04-21

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  13. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  14. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  15. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    Energy Technology Data Exchange (ETDEWEB)

    Mertyurek, Ugur, E-mail: mertyureku@ornl.gov; Gauld, Ian C., E-mail: gauldi@ornl.gov

    2016-02-15

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  16. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  17. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

  18. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  19. Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

    OpenAIRE

    Shelley, A.; 久語 輝彦; 嶋田 昭一郎; 大久保 努; 岩村 公道

    2004-01-01

    Neutronic study has been done for a PWR-type reduced-moderation water reactor with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void coefficient and a high burnup by using a MOX fuel. The results of the precise assembly burnup calculations show that the recommended numbers of seed and blanket layers are 15(S15) and 5(B5), respectively. By the optimization of axial configuration, the S15B5 assembly with the seed of 1000times2 mm high, internal blanket of 150 mm h...

  20. Current Status of Periodic Safety Review of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin; Ahn, Guk-Hoon; Lee, Choong Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A PSR for a research reactor became a legal requirement as the Nuclear Safety Act was amended and came into effect in 2014. This paper describes the current status and methodology of the first Periodic Safety Review (PSR) of HANARO that is being performed. The legal requirements, work plan, and process of implementing a PSR are described. Because this is the first PSR for a research reactor, it is our understating that the operating organization and regulatory body should communicate well with each other to complete the PSR in a timely manner. The first PSR of HANARO is under way. In order to achieve a successful result, activities of the operation organization such as scheduling, maintaining consistency in input data for review, and reviewing the PSR reports that will require intensive resources should be well planned. This means the operating organization needs to incorporate appropriate measures to ensure the transfer of knowledge and expertise arising from the PSR via a contractor to the operation organization. It is desirable for the Regulatory Body to be involved in all stage of the PSR to prevent any waste of resources and minimize the potential for a reworking of the PSR and the need for an additional assessment and review as recommended by foreign experts.

  1. Reduction of the pool-top radiation level in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong-Sung; Park, Sang-Jun; Kim, Heonil; Park, Yong-Chul; Choi, Young-San [HANARO Center, Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    HANARO is an open-tank-in-pool type reactor. Pool water is the only shielding to minimize the pool top radiation level. During the power ascension test of HANARO, the measured pool top radiation level was higher than the design value because some of the activation products in the coolant reached the pool surface. In order to suppress this rising coolant, the hot water layer system (HWL) was designed and installed to maintain l.2 meter-deep hot water layer whose temperature is 5degC higher than that of the underneath pool surface. After the installation of the HWL system, however, the radiation level of the pool-top did not satisfy the design value. The operation modes of the hot water layer system and the other systems in the reactor pool, which had an effect on the formation of the hot water layer, were changed to reduce pool-top radiation level. After the above efforts, the temperature and the radioactivity distribution in the pool was measured to confirm whether this system blocked the rising coolant. The radiation level at the pool-top was significantly reduced below one tenth of that before installing the HWL and satisfied the design value. It was also confirmed by calculation that this hot water layer system would significantly reduce the release of fission gases to the reactor hall and the environment during the hypothetical accident as well. (author)

  2. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G., E-mail: aabarca@isirym.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@iqn.upv.es [Universitat Politecnica de Valencia, (ISIRYM/UPV), (Spain). Institute for Industrial, Radiophysical and Environmental Safety; Concejal, A.; Melara, J.; Albendea, M., E-mail: acbe@iberdrola.es, E-mail: jls@iberdrola.es, E-mail: manuel.albendea@iberdrola.es [Iberdrola, Madrid (Spain); Soler, A., E-mail: asoler@iberdrola.es [SEA Propulsion SL, Madrid (Spain)

    2013-07-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  3. Transport of fresh MOX fuel assemblies for the MONJU initial core

    Energy Technology Data Exchange (ETDEWEB)

    Miura, Y.; Ouchi, Y.; Kurakami, J. [Power Reactor and Nuclear Fuel Development Corp., Tokai Works, Naka, Ibaraki (Japan); Usami, M. [Power Reactor and Nuclear Fuel Development Corp., Head Office Sankaido Building, Minato, Tokyo (Japan)

    1997-12-31

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author).

  4. Comparison of prediction models for Cherenkov light emissions from nuclear fuel assemblies

    Science.gov (United States)

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

    2017-06-01

    The Digital Cherenkov Viewing Device (DCVD) [1] is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the Cherenkov light produced by the assembly. Verifying that no rods have been substituted in the fuel, so-called partial-defect verification, is done by comparing the intensity measured with a DCVD with a predicted intensity, based on operator fuel declaration. The prediction model currently used by inspectors is based on simulations of Cherenkov light production in a BWR 8x8 geometry. This work investigates prediction models based on simulated Cherenkov light production in a BWR 8x8 and a PWR 17x17 assembly, as well as a simplified model based on a single rod in water. Cherenkov light caused by both fission product gamma and beta decays was considered. The simulations reveal that there are systematic differences between the model used by safeguards inspectors and the models described in this publication, most noticeably with respect to the fuel assembly cooling time. Consequently, if the intensity predictions are based on another fuel type than the fuel type being measured, a systematic bias in intensity with respect to burnup and cooling time is introduced. While a simplified model may be accurate enough for a set of fuel assemblies with nearly identical cooling times, the prediction models may differ systematically by up to 18 % for fuels with more varied cooling times. Accordingly, these investigations indicate that the currently used model may need to be exchanged with a set of more detailed, fuel-type specific models, in order minimize the model dependent systematic deviations.

  5. Status of Neutron Irradiation of Non-Nuclear Materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee-Nam; Cho, Man-Soon; Shin, Yoon-Taek; Park, Seng-Jae; Kang, Young Hwan; Jun, Byung-Hyuk; Kim, Chan-Joong; Park, Sang-Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The irradiation facilities have been mostly utilized for the KAERI nuclear research projects relevant to a commercial nuclear power reactor such as the ageing management and safety evaluation of the components. Based on the accumulated experience, HANARO has recently supported national R and D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. As neutron irradiation affects the structure of a material, radiation induced modification of materials has become a perspective method for the purposeful changes in material properties. Some irradiation tests of electro-magnetic materials were also performed at HANARO for scientific research of universities and the demand for neutron irradiation of the materials is increasing rapidly. Another research reactor that will specialize in radio-isotope and neutron transmutation doping (NTD) Si production and the demonstration of reactor design is under construction in Korea. Therefore, HANARO will specialize more on irradiation research. Based on its accumulated irradiation experience, HANARO has recently started new support of R and D relevant to the irradiation of electro-magnetic materials. In this paper, the status of utilization of the HANARO irradiation facilities for non-nuclear materials and the possibility of researches on new electro-magnetic materials using neutron irradiation are surveyed to encourage the utilization of HANARO. HANARO irradiation facilities have been actively utilized for various material irradiation tests requested by users. Although most irradiation tests have been related to national R and D relevant to nuclear power, demand for neutron irradiation of electro-magnetic materials is rapidly increasing at HANARO. Another research reactor, the KIJANG research reactor (KJRR), is under construction in Korea. New irradiation facilities including Neutron Transmutation Doping (NTD) facilities for power

  6. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  7. Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D.; Yang, Z.; Zhong, Y.; Xiao, Y.; Hu, L. [Chongqing Univ. (China). Key Lab. of Low-grade Energy Utilization Technologies and Systems

    2016-07-15

    In pressurized water reactors (PWRs), the spacer grids of the fuel assembly has significant impact on the thermal-hydraulic performance of the fuel assembly. Particularly, the spacer grids with the mixing vanes can dramatically enhance the secondary flow and have significant effect on the void distribution in the fuel assembly. In this paper, the CFD study has been carried out to analyze the effects of the spacer grid with the steel contacts, dimples and mixing vanes on the boiling two-phase flow characteristics, such as the two-phase flow field, the void distribution, and so on. Considered the influence of the boiling phase change on two-phase flow, a boiling model was proposed and applied in the CFD simulation by using the UDF (User Defined Function) method. Furthermore, in order to analyze the effects of the spacer grid with mixing vanes, the adiabatic (without boiling) two-phase flow has also been investigated as comparison with the boiling two-phase flow in the fuel assembly with spacer grids. The CFD simulation on two-phase flow in the fuel assembly with the proposed boiling model can predict the characteristics of two-phase flow better.

  8. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  9. Assaying Used Nuclear Fuel Assemblies Using Lead Slowing-Down Spectroscopy and Singular Value Decomposition

    Science.gov (United States)

    Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Chris J.; Warren, Glen A.

    2013-04-01

    This study investigates the use of a Lead Slowing-Down Spectrometer (LSDS) for the direct and independent measurement of fissile isotopes in light-water nuclear reactor fuel assemblies. The current study applies MCNPX, a Monte Carlo radiation transport code, to simulate the measurement of the assay of the used nuclear fuel assemblies in the LSDS. An empirical model has been developed based on the calibration of the LSDS to responses generated from the simulated assay of six well-characterized fuel assemblies. The effects of self-shielding are taken into account by using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the self-shielding functions from the assay of assemblies in the calibration set. The performance of the empirical algorithm was tested on version 1 of the Next-Generation Safeguards Initiative (NGSI) used fuel library consisting of 64 assemblies, as well as a set of 27 diversion assemblies, both of which were developed by Los Alamos National Laboratory. The potential for direct and independent assay of the sum of the masses of Pu-239 and Pu-241 to within 2%, on average, has been demonstrated.

  10. Transport of fresh MOX fuel assemblies for the Monju initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kurakami, J.; Ouchi, Y. [Power Reactor and Nuclear Fuel Development Corp., Tokai Works (Japan); Usami, M. [Power Reactor and Nuclear Fuel Development Corp., Tokyo Head Office, Power Reactor and Nuclear Fuel Development Corp., Ibaraki (Japan)

    1997-12-01

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author).

  11. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  12. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  13. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  14. Technical Requirements for Fabrication and Installation of Removable Shield for CNRF in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Shin, Jin Won

    2008-04-15

    This report details the technical requirements for the fabrication and installation of the removable shield for the Cold Neutron Research Facility (CNRF) in HANARO reactor hall. The removable shield is classified as non-nuclear safety (NNS), seismic category II, and quality class T. The main function of the removable shield is to do the biological shielding of neutrons and gamma from the CN port and the guides. The removable shield consists of block type walls and roofs that can be necessarily assembled, disassembled and moveable. These will be installed between the reactor pool wall and the CNS guide bunker in. This report describes technical requirements for the removable shield such as quality assurance, seismic analysis requirements, configuration, concrete compositions, fabrication and installation requirements, test and inspection, shipping, delivery, etc. Appendix is the technical specification of structural design and analysis. Attachments are composed of the technical specification for the fabrication of the removable shield, shielding design drawings and procurement quality requirements. These technical requirements will be provided to a contract for the manufacturing and installation.

  15. Investigation for the Fossil Embryo using Neutron Tomography at HANARO, KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Joo; Sim, Cheul Muu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Hee [National Science Museum, Daejeon (Korea, Republic of); Grellet-Tinner, Gerald [The Field Museum, Chicago (United States)

    2012-05-15

    Neutron imaging technique is one of non-destructive method. It is similar to X-ray and g-ray methods in using the different attenuation characteristics depending on materials. However, there is great difference between them. The mass attenuation coefficients of X-ray and g- ray monotonically increase with the atomic number since they interact with electrons. Thus X-ray image method does not supply sufficient contrast between similar atomic numbers. On the other hand, that of thermal neutrons depends much on the nucleus not electrons. Especially thermal neutrons easily penetrate most of metals, while they are attenuated well by such materials as hydrogen, water, boron, gadolinium and cadmium. Because of these unique characteristics of neutron, neutron imaging technique has been utilized for NDT or researches for next power sources (fuel cell or Li-Ion battery). Recently, dinosaur egg was found at the Aptian. Albian Algui Ulaan Tsav site, Mongolia. In this study, we applied the neutron imaging technique to investigate dinosaur embryo at Neutron Radiography Facility of HANARO, KAERI

  16. Development of a heat transfer correlation for the HPLWR fuel assembly by means of CFD analyses

    Energy Technology Data Exchange (ETDEWEB)

    Lycklama a Nijeholt, J.A.; Visser, D.C. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Laurien, E. [Univ. of Stuttgart, Stuttgart (Germany); Anglart, H. [Royal Inst. of Tech., Stockholm (Sweden); Chandra, L. [Indian Inst. of Tech., Rajasthan (India)

    2011-07-01

    The High Performance Light Water Reactor (HPLWR) has been under development in the HPLWR phase-2 project funded by the European Union. The HPLWR project started September 2006 and ended February 2010. Work package 5 within this project involves the improved understanding of heat transfer, CFD model development and validation, and the prediction of the heat transfer rate in a HPLWR fuel assembly. USTUTT, KTH, NRG and FZK contributed to this work package. The overall objective of work package 5 was the development of a heat transfer correlation for the prediction of the heat transfer rate in the HPLWR fuel assembly by means of CFD analyses. In the HPLWR fuel assembly, a helical wire has been selected as spacer and mixing device. This wire-wrap imposed a significant challenge in the development of the geometrical models for the CFD analyses. Due to the wire-wrap it was not possible to model a full fuel assembly consisting of 40 rods. Therefore, an alternative procedure has been adopted to develop a heat transfer correlation for the HPLWR fuel assembly. This procedure involved the definition of correction factors accounting for the effect of the rod bundle geometry and the wire-wrap spacer with respect to a smooth circular tube with super-critical water. The present paper describes the procedure followed in work package 5 of the HPLWR phase-2 project for the development of a heat transfer correlation for the HPLWR fuel assembly design and presents the derivation of the applied correction factors from a large set of CFD analyses for different representative geometries like an annulus, a single sub-channel and a 4 rod-bundle, all with and without inclusion of the wire wrap. (author)

  17. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  18. Process for recycling components of a PEM fuel cell membrane electrode assembly

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  19. Development of Out-pile Test Technology for Fuel Assembly Performance Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2007-03-15

    Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

  20. Fabrication of nuclear fuel assemblies in Mexico; Fabricacion de ensambles de combustible nuclear en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Medrano B, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: amb@nuclear.inin.mx

    2007-07-01

    In the Pilot Production Plant of Nuclear Fuel facilities (PPFCN) located in the Nuclear Center of Mexico; its were processed approximately 1000 Kg of powder of uranium dioxide with 11 different enrichments from 0.71 up to 3.95% U-235, the pellets were encapsulated in Zircaloy tubes and armed around 300 rods of nuclear fuel for to manufacture four assembles of nuclear fuel and a DUMMY for the qualification of processes, personnel and equipment. The project beginning in 1990 with the one agreement among General Electric, Federal Commission of Electricity (CFE) and the National Institute of Nuclear Research (ININ), after building the PPFCN, to install equipment, to design the parameters of production and to qualify us as suppliers of nuclear fuel; it was begins in 1994 the production of four GE9B assemblies that surrendered to the CNLV in May, 1996. In 1998 its were loaded in the unit 1 of the CNLV, assemble them of nuclear fuel with serial numbers INI002, INI003, INI004 and INI005 with an average enrichment of 3.03% U-235, four complete operational cycles worked including the central control cell. During the works of the ninth recharge of the unit 1 of the CNLV, September 20, 2002 were removed these assemblies from the reactor core reaching a burnt of 35313 MWD/TMU. (Author)

  1. Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

    Science.gov (United States)

    Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

    2017-01-03

    Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm 2 ) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

  2. Feasibility study on the verification of fresh fuel assemblies in shipping containers

    Energy Technology Data Exchange (ETDEWEB)

    Swinth, K.L.; Tanner, J.E.

    1990-09-01

    The purpose of this study was to examine the feasibility of using various nondestructive measurement techniques to determine the presence of fuel assemblies inside shipping containers and to examine the feasibility of measuring the fissile content of the containers. Passive and active techniques based on both gamma and neutron assay were examined. In addition, some experiments and calculations were performed to evaluate neutron techniques. Passive counting of the 186 keV gamma from {sup 235}U is recommended for use as an attributes measurement technique. Experiments and studies indicated that a bismuth germanate (BGO) scintillator is the preferred detector. A properly designed system based on this detector will provide a compact detector that can selectively verify fuel assemblies within a shipping container while the container is in a stack of similarly loaded containers. Missing fuel assemblies will be readily detected, but gamma counting of assemblies cannot detect changes in the fissile content of the inner rods in an assembly. If a variables technique is required, it is recommended that more extensive calculations be performed and removal of the outer shipping container be considered. Marking (sealing) of the assemblies with a uniquely identifiable transponder was also considered. This would require the development of procedures that would assure proper application and removal of the seal. When change to a metal outer container occurs, the technique will no longer be useful unless a radiolucent window is included in the container. 20 refs., 7 figs., 2 tabs.

  3. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  4. Layer-by-layer self-assembled carbon nanotube electrode for microbial fuel cells application.

    Science.gov (United States)

    Roh, Sung-Hee

    2013-06-01

    Anodic material is usually a limiting factor in power generation in microbial fuel cell (MFC). A meditatorless two-chambered MFC was constructed using new type of carbon nanotube (CNT)-based electrode architecture by layer-by-layer (LBL) self-assembly to increase power density. Multi-walled carbon nanotube (MWNT) and polyeletrolyte polyethyleneimine (PEI) were employed to modify carbon paper (CP) electrode utilizing a LBL self-assembly technique, and the electrochemical properties and performance of the modified electrode as an anode in MFC were investigated. The CNT-based LBL self-assembled electrode showed a better electrochemical performance than that of unmodified CP electrode. The self-assembled MWNT/PEI onto CP produced the maximum power density of 480 mW/m2, which was 48% larger than that of the plain CP anode. The CNT-based LBL self-assembled electrode therefore offers good prospects for application in MFCs.

  5. A monolithic silicon-based membrane-electrode assembly for micro fuel cells

    Science.gov (United States)

    Yuzova, V. A.; Merkushev, F. F.; Semenova, O. V.

    2017-08-01

    We report the basic possibility of creating a micro fuel cell (MFC) with a monolithic silicon-based membrane-electrode assembly (MEA), which employs a porous three-layer framework structure manufactured by two-sided anodic etching of a 500-μm-thick silicon wafer. A technology of MEAs for MFCs is described.

  6. RELIABILITY of FUEL ASSEMBLY EFFLUENT TEMPERATURES UNDER L0CA/LOPA CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Sachs, A.D.

    1999-06-21

    The purpose of this study was to ascertain whether or not the K-Reactor safety computers could calculate primarily false positive, but also false negative, and ''on-scale'' misleading fuel assembly average effluent temperatures (AETs) due to relatively large temperature changes in or flooding of the -36 foot elevation isothermal box during a LOCA/LOPA.

  7. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    Energy Technology Data Exchange (ETDEWEB)

    Lomax, F.D. Jr.; James, B.D. [Directed Technologies, Inc., Arlington, VA (United States); Mooradian, R.P. [Ford Motor Co., Dearborn, MI (United States)

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  8. CFD Analysis of Turbulent Flow in Nose Piece for SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; In, Wang Kee; Yoon, Kyoung Ho; Cheon, Jin Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A CFD analysis was performed to simulate coolant flow in nose piece of SFR fuel assembly. The unsteady CFD simulation was conducted using RANS turbulence models and LES. The CFD prediction shows a violent swirling and recirculating flow inside nose piece due to staggered arrangement of side orifices. The lateral and axial flows generate a complex unsteady three-dimensional flow in the nose piece. A higher vorticity was predicted to occur inside the nose piece particularly by the LES. The loss coefficient of the side orifice is estimated to be 2.5-3.0 which is somewhat lower than the experimental value of 3.5. The fuel assembly for sodium-cooled fast reactor(SFR) consists of nose piece, lower/upper reflector, fuel bundle and lifting lug. The nose piece is an inlet nozzle with nine(9) side orifices. The coolant from core inlet plenum is forced to flow inside the nose piece through side orifices and directed upward to fuel bundle. Hence, the coolant flow in nose piece is highly unsteady turbulent flow and complex flow due to staggered side orifices. This paper presents the CFD(computational fluid dynamics) predictions of turbulent flow in nose piece and pressure loss coefficient of side orifices for SFR fuel assembly.

  9. Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2016-12-15

    Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

  10. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  11. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  12. Burnable Absorbers with Enriched Er-167 in PWR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choe, Jiwon; Kong, Chidong; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Many advanced PWRs are required to have a 24-month operating cycle to improve plant economy, and to keep the boron concentration low to allow an adequately negative moderator feedback during any ATWS event through 100% core life. Unfortunately, longer cycles require higher uranium-235 enrichment and initial boron concentration in the reactor coolant. The amount of soluble boron is limited due to the requirement that the MTC must remain negative over the fuel cycle. Too much boron, typically greater than 1,300 ppm at full power, will make the MTC positive. The optimal design of burnable absorbers is key to the feasibility of this extended cycle and low boron core below the design limit of peak pin power. New concepts for burnable absorbers include changing the materials and geometry in the burnable absorber. k{sub inf}, peaking factor, MTC, and control rod worth of new BAs were compared with those of the conventional BA. A new enriched Er-167 based BA has been proposed and, from three test cases, it was shown that the Erbium burnable absorber is favorable to counterbalance the power peak and Gadolinium burnable absorber is favorable to flattening k{sub inf} trends over burnup.

  13. Seismic response Analyses of Hanaro in-chimney bracket structures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Ryu, J.S.; Cho, Y.G.; Lee, H.Y.; Kim, J.B

    1999-05-01

    The in-chimney bracket will be installed in the upper part of chimney, which holds the capsule extension pipes in upper one-third of length. For evaluating the effects on the capsules and related reactor structures, ANSYS finite element analysis model is developed and the dynamic characteristics are analyzed. The seismic response anlayses of in-chimney bracket and related reactor structures of HANARO under the design earthquake response spectrum loads of OBE (0.1 g) and SSE (0.2 g) are performed. The maximum horizontal displacements of the flow tubes are within the minimum half gaps between close flow tubes, it is expected that these displacement will not produce any contact between neighbor flow tubes. The stress values in main points of reactor structures and in-chimney bracket for the seismic loads are also within the ASME Code limits. It is also confirmed that the fatigue usage factor is muchless than 1.0. So, any damage on structural integrity is not expected when an in-chimney bracket is installed to upper part of the reactor chimney. (author). 12 refs., 24 tabs., 37 figs.

  14. Improvement of the reactivity computer for HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jin; Park, S. J.; Jung, H. S.; Choi, Y. S.; Lee, K. H.; Seo, S. G

    2001-04-01

    The reactivity computer in HANARO has a dedicated neutron detection system for experiments and measurements of the reactor characteristics. This system consists of a personal computer and a multi-function I/O board, and collects the signals from the various neutron detectors. The existing hardware and software are developed under the DOS environment so that they are very inconvenient to use and have difficulties in finding replacement parts. Since the continuity of the signal is often lost when we process the wide rang signal, the need for its improvement has been an issue. The purpose of this project is to upgrade the hardware and software for data collection and processing in order for them to be compatible with Windows{sup TM} operating system and to solve the known issue. We have replaced the existing system with new multi-function I/O board and Pentium III class PC, and the application program for the wide range reactivity measurement and multi-function signal counter have been developed. The newly replaced multi-function I/O board has seven times fast A/D conversion rate and collects sufficient amount of data in a short time. The new application program is user-friendly and provides various useful information on its display screen so that the ability of data processing and storage has been very much enhanced.

  15. Thermal–hydraulic numerical simulation of fuel sub-assembly using a dedicated meshing tool

    Energy Technology Data Exchange (ETDEWEB)

    Cadiou, Thierry, E-mail: thierry.cadiou@cea.fr; Saxena, Aakanksha

    2015-12-15

    As the CEA is involved in the pre-conceptual design phase of a Sodium-cooled Fast Reactor (SFR), the thermal–hydraulics modeling of sodium flow in the reactor core is a key scientific subject, not only due to the innovations proposed for the core design but also for the cost and the difficulties encountered to carry out experiments with sodium. Taking advantage of the progress made in numerical simulation and associated computational time, the sodium flow in a fuel pin sub-assembly is characterized in this work. The fuel pin sub-assembly is composed of 217 fuel pins, each wrapped by spacer wire, and surrounded by a hexagonal tube. In order to overcome the main limitation on the required number of mesh cells for modeling such a complex geometry, an original meshing tool, developed for this purpose, was necessary and is presented in this paper. This approach reveals to be of great help for modeling the sodium flow. Indeed, the pressure drop in the rod bundle is firstly evaluated. In addition, the local effects (at the scale of fuel pin) and global effects (at the scale of fuel pin bundle) for sodium velocity and temperature gradients for the alleged homogenization made by the spacer wire on the sodium flow are examined and clarified. Taking into account these results, the optimization of the fuel bundle geometry can be considered in order to homogenize the outlet temperature distribution in nominal condition. Lastly, the definition and the analysis of experiments on sub-assembly to be carried out in future experimental CEA platform, will also take advantage of this approach.

  16. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Song, K. C. [and others

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

  17. Effect of nonuniform 3-D void distribution within BWR fuel assemblies on neutronic characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu; Hyoudou, Hideaki; Kitada, Takanori [Osaka Univ., Suita (Japan). Dept. of Nuclear Engineering]. E-mail: takeda@nucl.eng.osaka-u.ac.jp; Kosaka, Shinya; Ikeda, Hideaki [Tepco Systems Corp., Tokyo (Japan)]. E-mail: kisaka-shinya@tepsys.co.jp

    2003-07-01

    The effect of nonuniform distribution of void fractions within fuel assemblies of BWR cores on the neutronic characteristics has been evaluated by coupling the thermal hydraulic and neutronic calculations. It is found that the effect is large for assemblies with Gd rods because of the small void fraction for subchannels with Gd rods. The axially averaged K{sub {infinity}} is reduced by about 0.1 {approx} 0.3% {delta}k/k compared with the case with uniform void distribution. The power peaking is decreased by 0.6 {approx} 1.3%. The reason of these changes is discussed. (author)

  18. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  19. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    Science.gov (United States)

    Forsberg, Charles W.

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

  20. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology

    2012-11-15

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

  1. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    Science.gov (United States)

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  2. Calculated neutron-source spectra from selected irradiated PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Bosler, G.E.; Phillips, J.R.

    1981-12-01

    The energy spectra of neutrons emitted from a pressurized-water-reactor fuel assembly have been calculated for a variety of exposures and cooling times. They are presented in graphical form. Some effects of initial enrichment are also included. Neutrons from spontaneous fissions were given either a Maxwellian temperature of 1.2 or 1.5 MeV, depending on whether they were due to plutonium and uranium nuclides or curium nuclides. A single (..cap alpha..,n) spectrum was deemed sufficient to represent the neutrons from all the alpha-emitting nuclides. The proportions of the nuclides undergoing spontaneous fission and those emitting alpha particles were determined from calculated atom densities. The particular pressurized-water-reactor fuel assembly assumed for this purpose was of the type used in the H.B. Robinson Unit-2 power plant (740 MWe).

  3. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-06-04

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium.

  4. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  5. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2008-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  6. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  7. Vibration test report for in-chimney bracket and instrumented fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, D. B.; Cho, Y. G.; Ahn, G. H.; Lee, J. H.; Park, J.H

    2000-10-01

    The vibration levels of in-chimney bracket structure which is installed in reactor chimney and instrumented fuel assembly(Type-B Bundle) are investigated under the steady state normal operating condition of the reactor. For this purpose, 4 acceleration data on the guide tube of the instrumented fuel assembly and in-chimney bracket structures subjected to fluid induced vibration are measured. For the analysis of the vibration data, vibration analysis program which can perform basic time and frequency domain analysis, is prepared, and its reliability is verified by comparing the analysis results with those of commercial analysis program(I-DEAS). In time domain analysis, maximum amplitudes, and RMS values of accelerations and displacements from the measured vibration signal, are obtained. The frequency components of the vibration data are analyzed by using the frequency domain analysis. These analysis results show that the levels of the measured vibrations are within the allowable level, and the low frequency component near 10 Hz is dominant in the vibration signal. For the evaluation of the structural integrity on the in-chimney bracket and related structures including the instrumented fuel assembly, the static analysis for ANSYS finite element model is carried out. These analysis results show that the maximum stresses are within the allowable stresses of the ASME code, and the maximum displacement of the top of the flow tube is within the displacement limit. Therefore any damage on the structural integrity is not expected when the irradiation test is performed using the in-chimney bracket.

  8. Automated assembling of single fuel cell units for use in a fuel cell stack

    Science.gov (United States)

    Jalba, C. K.; Muminovic, A.; Barz, C.; Nasui, V.

    2017-05-01

    The manufacturing of PEMFC stacks (POLYMER ELEKTROLYT MEMBRAN Fuel Cell) is nowadays still done by hand. Over hundreds of identical single components have to be placed accurate together for the construction of a fuel cell stack. Beside logistic problems, higher total costs and disadvantages in weight the high number of components produce a higher statistic interference because of faulty erection or material defects and summation of manufacturing tolerances. The saving of costs is about 20 - 25 %. Furthermore, the total weight of the fuel cells will be reduced because of a new sealing technology. Overall a one minute cycle time has to be aimed per cell at the manufacturing of these single components. The change of the existing sealing concept to a bonded sealing is one of the important requisites to get an automated manufacturing of single cell units. One of the important steps for an automated gluing process is the checking of the glue application by using of an image processing system. After bonding the single fuel cell the sealing and electrical function can be checked, so that only functional and high qualitative cells can get into further manufacturing processes.

  9. Influence of moderator to fuel ratio (MFR) on burning thorium in a subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Wojciechowski, Andrzej, E-mail: andrzej.wojciechowski@ncbj.gov.pl [National Center for Nuclear Research, Otwock-Swierk (Poland); Joint Institute for Nuclear Research, Dubna (Russian Federation)

    2014-10-15

    The conversion ratio (CR) of Th-232 to U-233 calculation results for a subcritical reactor assembly is presented as a function of MFR, burnup, power density (PD) and fissile concentration. The calculated model is based on subcritical assembly which makes configuration of fuel rods and volumes of moderator and coolant changes possible. This comfortable assembly enables investigation of CR in a thorium cycle for different value of MFR. Additionally, the calculation results of U-233 saturation concentration are explained by mathematical model. The value of MFR main influences the saturation concentration of U-233 and fissile and the fissile concentration dependence of CR. The saturation value of CR is included in the range CR ∈ (0.911, 0.966) and is a slowly increasing function of MFR. The calculations were done with a MCNPX 2.7 code.

  10. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  11. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  12. Advanced manufacturing of intermediate temperature, direct methane oxidation membrane electrode assemblies for durable solid oxide fuel cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ITN proposes to create an innovative anode supported membrane electrode assembly (MEA) for solid oxide fuel cells (SOFCs) that is capable of long-term operation at...

  13. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  14. The status of the safeguards implementation under the State-Level Approach at the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. S.; Lee, B. D.; Kim, I. C.; Kim, H. J.; Jung, J. A.; Lee, S. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The IAEA developed the SLA(State-Level Approach) for the States in order to maximize effectiveness of safeguards in an environment of constrained resources. The SLA has been implemented at KAERI-Daejeon site in the ROK since 2015. The ten nuclear facilities and one LOF(Location Outsides Facility) of the KAERI-Daejeon site are grouped into three categories under the SLA. The HANARO(High flux Advanced Neutron Application ReactOr) and PIEF(Post Irradiation Examination Facility) are involved in the category I “self-contained capability” facilities that have at least one significant quantity of suitable nuclear material and which could support undeclared plutonium production/separation activities without other supporting infrastructures. This paper described the status of the safeguards implementation at the HANARO involved in the category I under the SLA. The status of a model inventory management system for a research reactor developed in 2013 was also investigated. In this paper, the features and status of the safeguards implementation of the HANARO under the SLA were analyzed. Under the SLA, the monthly, quarterly and annual advanced facility operational information for the HANARO has been submitted to the IAEA in a timely manner. The IAEA inspection at HANARO has been successfully performed under the SLA. It is expected that the safeguards implementation work at HANARO under the SLA has the similar level with that under IS. Under the SLA, the data occurred from the surveillance cameras and other equipment installed at HANARO enables to transmit remotely to the IAEA. The IAEA is targeting 2017~2018 to upgrade them. In addition, the development status of a model inventory management system for a research reactor was investigated. It aims at controlling the material inventory for the nuclear material accounting work and the convenient facility operation. The major functions of it are to trace the transfer history of the nuclear materials and non-nuclear materials

  15. Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

    Science.gov (United States)

    Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

    1985-01-01

    A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

  16. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Ama, Tsuyoshi; Hyoudou, Hideaki; Takeda, Toshikazu [Osaka Univ., Graduate School of Engineering, Suita, Osaka (Japan); Ikeda, Hideaki; Kosaka, Shinya [Tepco Systems Corp., In-core Fuel Management Department, Tokyo (Japan)

    2002-05-01

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  17. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  18. Shielding Analysis for the Lower Reflector Block of a PGSFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae-Yong; Hartanto, Donny [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The PGSFR uses single enriched uranium metal-alloy fuel, U-10%Zr, initially and it will be gradually converted into a TRU metal-alloy (UTRU- 10%Zr) fueled core. The fuel assemblies are arranged into a 7-hexagonal-ring configuration, surrounded by stainless steel reflectors and B{sub 4}C shields. Each fuel assembly consists of a fuel rod bundle with 217 fuel pins and a lower/upper reflector block. A lower and an upper reflector blocks are installed inside the fuel assembly to reduce the radiation damage on the core support plate and the radioactivity on head access area respectively. This design can satisfy the radiation damage limit on permanent structure, but it has a high coolant pressure drop caused by the warped coolant path when analyzed using detailed CFD. To address this issue, three reflector block designs which have a lower pressure drop were proposed. In this paper, we analyzed the shielding capability of the three new reflector blocks. The shielding evaluations were performed for the new different reflector block models. In order to describe the complex internal/external structure of the reflector design in detailed, Serpent 2 code was employed instead of MCNP6. The code comparison between MCNP6 and Serpent 2 was examined and confirmed good agreement within a few percent relative error. By the 3-D whole core modeling, the DPA analyses for 3 new reflector block models were conducted and all models satisfied the DPA limit on SS316 support grid plate. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis.

  19. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  20. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  1. Controlling the hydrophilicity and contact resistance of fuel cell bipolar plate surfaces using layered nanoparticle assembly

    Science.gov (United States)

    Wang, Feng

    Hybrid nanostructured coatings exhibiting the combined properties of electrical conductivity and surface hydrophilicity were obtained by using Layer-by-Layer (LBL) assembly of cationic polymer, silica nanospheres, and carbon nanoplatelets. This work demonstrates that by controlling the nanoparticle zeta (zeta) potential through the suspension parameters (pH, organic solvent type and amount, and ionic content) as well as the assembly sequence, the nanostructure and composition of the coatings may be adjusted to optimize the desired properties. Two types of silica nanospheres were evaluated as the hydrophilic component: X-TecRTM 3408 from Nano-X Corporation, with a diameter of about 20 nm, and polishing silica from Electron Microscopy Supply, with diameter of about 65 nm. Graphite nanoplatelets with a thickness of 5~10nm (Aquadag RTM E from Acheson Industries) were used as electrically conductive filler. A cationic copolymer of acrylamide and a quaternary ammonium salt (SuperflocRTM C442 from Cytec Corporation) was used as the binder for the negatively charged nanoparticles. Coatings were applied to gold-coated stainless steel substrates presently used a bipolar plate material for proton exchange membrane (PEM) fuel cells. Coating thickness was found to vary nearly linearly with the number of polymer-nanoparticle layers deposited while a monotonic increase in coating contact resistance was observed for all heterogeneous and pure silica coatings. Thickness increased if the difference in the oppositely charged zeta potentials of the adsorbing components was enhanced through alcohol addition. Interestingly, an opposite effect was observed if the zeta potential difference was increased through pH variation. This previously undocumented difference in adsorption behavior is herein related to changes to the surface chemical heterogeneity of the nanoparticles. Coating contact resistance and surface wettability were found to have a more subtle dependence on the assembly

  2. Impedance Analysis of the Conditioning of PBI–Based Electrode Membrane Assemblies for High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Vang, Jakob Rabjerg; Andreasen, Søren Juhl

    2013-01-01

    This work analyses the conditioning of single fuel cell assemblies based on different membrane electrode assembly (MEA) types, produced by different methods. The analysis was done by means of electrochemical impedance spectroscopy, and the changes in the fitted resistances of the all the tested...

  3. Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new.

  4. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  5. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Douglas Alan [Univ. of Florida, Gainesville, FL (United States)

    1991-01-01

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  6. Irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Song, K. C.; Park, H. S. and others

    2000-04-01

    The objective of the irradiation test of DUPIC fuel at HANARO is to obtain the data of in-core behavior and evaluate the nuclear, thermal and mechanical performance of DUPIC fuel. The irradiation of DUPIC fuel will start at April 25, 2000 for about 2 months, and the burnup of 2,000 MWD/MTU will be attained for this period. The pre-irradiation examinations for DUPIC fuel, such as visual inspection, dimension measurement, He leak test and microstructure observation, was carried out. The post-irradiation examination items for the irradiated DUPIC fuel are planned to be the NDA test, visual inspection and dimension measurement, as well as the analyses for the fission gas release, the microstructure of pellets and the distribution and shape of imbedded nuclides. The DUPIC mini-elements were fabricated in the DFDF (IMEF M6 cell) using the G23-G2 rod. For the HANARO core calculation, the initial composition of DUPIC fuel was estimated using ORIGEN-2 code based on the burnup history of the G23-G2 rod. The design features of DUPIC pellets, the mini-element and the irradiation capsule, were supplemented considering the characteristics of DUPIC fuel and the results from the irradiation test of the simulated DUPIC fuel performed in 1999. The nuclear, thermohydraulic and mechanical characteristics of DUPIC fuel under the normal operation condition were evaluated for the safety analysis on the HANARO. Using these results, potential accidents initiated by DUPIC fuel were estimated, and Safety analyses on the locked rotor and RIA accidents were carried out in order to assess the integrity of DUPIC fuel under the accident condition initiated by the HANARO. Based on the results of these safety analyses, the supplemental countermeasures for securing the sufficient thermal margins were set up, as well. At the last, similar overseas and domestic cases were introduced.

  7. Non-fuel assembly components: 10 CFR 61.55 classification for waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Migliore, R.J.; Reid, B.D.; Fadeff, S.K.; Pauley, K.A.; Jenquin, U.P.

    1994-09-01

    This document reports the results of laboratory radionuclide measurements on a representative group of non-fuel assembly (NFA) components for the purposes of waste classification. This document also provides a methodology to estimate the radionuclide inventory of NFA components, including those located outside the fueled region of a nuclear reactor. These radionuclide estimates can then be used to determine the waste classification of NFA components for which there are no physical measurements. Previously, few radionuclide inventory measurements had been performed on NFA components. For this project, recommended scaling factors were selected for the ORIGEN2 computer code that result in conservative estimates of radionuclide concentrations in NFA components. These scaling factors were based upon experimental data obtained from the following NFA components: (1) a pressurized water reactor (PWR) burnable poison rod assembly, (2) a PVM rod cluster control assembly, and (3) a boiling water reactor cruciform control rod blade. As a whole, these components were found to be within Class C limits. Laboratory radionuclide measurements for these components are provided in detail.

  8. Impact characteristic of the peculiar spacer grids for advanced fuel assembly by drop type impact test

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Kyung Ho; Kim, Hyung Kyu; Kang, Heung Seok; Song, Kee Nam [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    The spacer grid is one of the main structural components in the fuel assembly, which supports the fuel rods, guides cooling water, and protects the system from an external impact load, such as earthquakes. The former is related to the mechanical integrity of the grid spring and dimple, and the latter is related to the structural integrity of the spacer grid. In this report, the drop type impact test is established that the mechanical strength of the candidate partial spacer grids for advanced PWR fuel assembly. It is accomplished by the comparison with the commercial same cell size spacer grid, which is composed of 5 x 5 partial V5H grid shape by Westinghouse design. Because of there is no specified design criterion about the structural strength of the partial spacer grid. As the results of them, two influential spacer grids are similar with that the reference grid. And the 5 x 5 cell type partial spacer grid is extended the 1/4 size model (8 x 8 cell type ) for verifying the mechanical performance, and then the negative points will be improving for detail design. 9 refs., 33 figs., 4 tabs. (Author)

  9. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Programa de Pós-Graduação em Ciências e Técnicas Nucleares

    2017-07-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k{sub ef{sub f}}. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi{sub L}ayer{sub P}erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  10. Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2017-01-01

    Full Text Available This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

  11. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  12. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  13. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  14. Stress and plastic deformation of MEA in fuel cells. Stresses generated during cell assembly

    Energy Technology Data Exchange (ETDEWEB)

    Bograchev, Daniil [Frumkin Institute of Physical Chemistry and Elecrtrochemistry RAN, Leninski prospekt 31, Moscow 117071 (Russian Federation); Gueguen, Mikael; Grandidier, Jean-Claude [Laboratoire de Physique et Mecanique des Materiaux, LMPM UMR CNRS 6617, ENSMA, Teleport 2, 1 av. Clement Ader, BP 40109 86962 Futuroscope Cedex (France); Martemianov, Serguei [Laboratoire d' Etudes Thermiques, LET UMR CNRS 6608, ESIP-University of Poitiers, 40 av. du Recteur Pineau, 86022 Poitiers (France)

    2008-05-15

    A linear elastic-plastic 2D model of fuel cell with hardening is developed for analysis of mechanical stresses in MEA arising in cell assembly procedure. The model includes the main components of real fuel cell (membrane, gas diffusion layers, graphite plates, and seal joints) and clamping elements (steel plates, bolts, nuts). The stress and plastic deformation in MEA are simulated with ABAQUS code taking into account the realistic clamping conditions. The stress distributions are obtained on the local and the global scales. The first one corresponds to the single tooth/channel structure. The global scale deals with features of the entire cell (the seal joint and the bolts). Experimental measurements of the residual membrane deformations have been provided at different bolts torques. The experimental data are in a good agreement with numerical predictions concerning the beginning of the plastic deformation. (author)

  15. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  16. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  17. Coupling of MCNPX with a sub-channel code for analysis of a HPLWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waata, C.; Schulenberg, T.; Cheng Xu [Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, (Germany)

    2005-07-01

    Full text of publication follows: The High Performance Light Water Reactor (HPLWR) project was launched in 2000 under the 5. Framework Program of the European Commission. The main objective of this project was to study the technical and economic feasibility of light water reactors operating at supercritical pressure conditions. This study aims to achieve high thermal efficiency of the nuclear power plant with operating conditions of pressure at 25 MPa, coolant temperature of about 510?C and an efficiency of up to 45%. The utilization of supercritical water as coolant and moderator in the HPLWR core introduces some challenges in the design of the HPLWR core due to the special behavior of the thermal-physical properties of water under super-critical pressure conditions. At supercritical pressure conditions, water does not exhibit a phase change. Therefore no boiling phenomenon occurs in the reactor core. However, there exist a strong variation in the water density in the core as the temperature changes across the pseudo-critical value. The strong variation in the water density affects strongly to the neutron-physical behavior in the core. Therefore, for an accurate and detailed design analysis of a HPLWR core, coupled analysis of neutron-physics with thermal-hydraulics is mandatory. Although extensive activities have been carried worldwide on the design of super-critical pressure light water reactors, accurate design analysis with neutron-physical/thermal-hydraulic coupling is still very limited. In the present study, the Monte-Carlo code, MCNPX, is coupled with the sub-channel analysis code, STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions), which was developed specifically for fuel assemblies of supercritical water cooled reactors and is also flexible for complex fuel assembly designs. In this paper, a short description about both codes is given. The coupling methodology and procedure is presented and assessed. A

  18. NUCLEAR DATA UNCERTAINTY PROPAGATION FOR A TYPICAL PWR FUEL ASSEMBLY WITH BURNUP

    Directory of Open Access Journals (Sweden)

    D. ROCHMAN

    2014-06-01

    Full Text Available The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling expert working group. The “Fast Total Monte Carlo” method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

  19. Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

    1985-04-21

    Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.

  20. Feasibility study on the transient fuel test loop installation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs.

  1. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ; Evaluacion termomecanica de los ensambles combustibles fabricados en el ININ

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  2. Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

    2017-01-15

    Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

  3. On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)]. E-mail: fabian.jatuff@psi.ch; Macku, K. [Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Chawla, R. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2006-01-15

    Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO{sub 2} fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.

  4. Electric power generation by a submersible microbial fuel cell equipped with a membrane electrode assembly

    DEFF Research Database (Denmark)

    Min, Booki; Poulsen, Finn Willy; Thygesen, Anders

    2012-01-01

    Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test, the maxi......Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test......, the maximum power density was 631mW/m2 at current density of 1772mA/m2 at 82Ω. With 180-Ω external resistance, one set of the electrodes on the same side could generate more power density of 832±4mW/m2 with current generation of 1923±4mA/m2. The anode, inclusive a biofilm behaved ohmic, whereas a Tafel type...... behavior was observed for the oxygen reduction. The various impedance contributions from electrodes, electrolyte and membrane were analyzed and identified by electrochemical impedance spectroscopy. Air flow rate to the cathode chamber affected microbial voltage generation, and higher power generation...

  5. A comparative study of MATRA-LMR/FB with CFD on a fuel assembly in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jin; Chang, Won-Pyo; Jeong, Jae-Ho; Ha, Kwi-Seok; Lee, Kwi-Lim; Lee, Seung Won; Choi, Chiwoong; Ahn, Sang-Jun [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Some of its models were modified to be eligible for the analysis of the SFR sub-channel blockage with the wire-wrapped pins. The wire-forcing-function used in the MATRA-LMR, which allocates a forced flow with an empirical correlation for the flow effect of the wire-wrap, was replaced with the Distributed Resistance Model. The Distributed Resistance Model has generally been believed to represent the effect more realistically than the wire-forcing-function. A semi-implicit numerical method was applied to resolve a flow reversal problem, which could not be handled by the former fully implicit method. A code-to-code comparison study was also performed as part of an effort to supplement the qualification. Although MATRA-LMR-FB was qualified based on available experimental data including a code-to-code comparative analysis, it was still hard to say that the level of confidence was enough to apply it to the SFR design with full satisfaction. Additional studies are therefore needed to supplement the qualification of MATRA-LMR-FB. In this study, a code-to-code comparative study was conducted as part of an effort to supplement the qualification of MATRA-LMR-FB. The comparison between MATRA-LMR-FB and the CFD code, CFX, was carried out on a 91-pin fuel assembly based on a 217 pin fuel assembly in a PGSFR to assess the MATRA-LMR-FB prediction capability.

  6. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  7. Development of HANARO human factors management plan and evaluation of BCS display

    Energy Technology Data Exchange (ETDEWEB)

    Oh, I. S.; Lee, J. W.; Lee, Y. H. [and others

    2004-01-01

    In this study, human factors evaluation of BCS display design was performed. We adopted the suitability of design elements of BCS display as human factors evaluation measure. And, we also adopted guideline based evaluation, field survey and expert evaluation as evaluation method. The checklist was utilized for the evaluation, and the results of evaluation were well arranged in the evaluation format. We did not find out the HED (Human Engineering Discrepancy) impede safety of HANARO, except some necessary items to improve during short periods. We also provide some items of improvement for the enhancement of safety and operator's performance in the aspect of long periods. If the proposed improvement items were completely fulfilled, the more improved safety of HANARO will be secured.

  8. Temperature monitoring using fibre optic sensors in a lead-bismuth eutectic cooled nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Acoustics and Vibration Research Group (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre, (SCK-CEN), Boeretang 200, Mol (Belgium); Lamberti, A.; Ertveldt, J.; Rezayat, A.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Acoustics and Vibration Research Group (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre, (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2016-02-15

    Highlights: • We demonstrate the use of optical fibre sensors in lead-bismuth cooled installations. • In this first of a kind experiment, we focus on temperature measurements of fuel rods • We acquire the surface temperature with a resolution of 30 mK. • We asses the condition of the installation during different steps of the operation. - Abstract: In-core temperature measurements are crucial to assess the condition of nuclear reactor components. The sensors that measure temperature must respond adequately in order, for example, to actuate safety systems that will mitigate the consequences of an undesired temperature excursion and to prevent component failure. This issue is exacerbated in new reactor designs that use liquid metals, such as for example a molten lead-bismuth eutectic, as coolant. Unlike water cooled reactors that need to operate at high pressure to raise the boiling point of water, liquid metal cooled reactors can operate at high temperatures whilst keeping the pressure at lower levels. In this paper we demonstrate the use of optical fibre sensors to measure the temperature distribution in a lead-bismuth eutectic cooled installation and we derive functional input e.g. the temperature control system or other systems that rely on accurate temperature actuation. This first-of-a-kind experiment demonstrates the potential of optical fibre based instrumentation in these environments. We focus on measuring the surface temperature of the individual fuel rods in the fuel assembly, but the technique can also be applied to other components or sections of the installation. We show that these surface temperatures can be experimentally measured with limited intervention on the fuel pin owing to the small geometry and fundamental properties of the optical fibres. The unique properties of the fibre sensors allowed acquiring the surface temperatures with a resolution of 30 mK. With these sensors, we assess the condition of the test section containing the fuel

  9. Development and characterization of proton conductive membranes and membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Jiang, Ruichun

    Polymer electrolyte membrane fuel cells (PEMFCs), including hydrogen fuel cells (PEMFCs) and direct methanol fuel cells (DMFCs), are considered as attractive electrical power sources. However, there are some technical obstacles that impede the commercialization of PEMFCs. For instance, in H 2-PEMFCs, carbon monoxide (CO) poisoning of the anode catalyst causes serious performance loss; in DMFCs, methanol crossover through the membrane reduces the overall fuel cell efficiency. This work focused on: (1) developing high performance membrane electrode assemblies (MEAs) and investigating their behavior at higher temperature H2-PEMFC with H2+CO as the fuel; (2) improving DMFCs efficiency by preparing low methanol crossover/good proton conductivity membranes based on NafionRTM matrix; (3) synthesizing and modifying low cost sulfonated hydrocarbon (SPEEK) membranes for both H2-PEMFCs and DMFCs applications. High performance membrane electrode assemblies (MEAs) with composite NafionRTM-TeflonRTM-Zr(HPO 4)2 membranes were prepared, optimized and characterized at higher temperature (> 100°C)/lower relative humidity (oxidation mechanism of H2/CO in higher temperature PEMFC was investigated and simulated. Two type of membranes based on NafionRTM matrix were prepared: silica/NafionRTM membrane and palladium impregnated NafionRTM (Pd-NafionRTM) membrane. The composite silica/NafionRTM membrane was developed by in-situ sol-gel reaction followed by solution casting, while the Pd-NafionRTM was fabricated via a supercritical fluid CO2 (scCO 2) route. Reduced methanol crossover and enhanced efficiency was observed by applying each of the two membranes to DMFCs. In addition, the research demonstrated that scCO2 is a promising technique for modifying membranes or depositing nano-particle electrocatalysts onto electrolyte. Sulfonated poly(ether ether ketone) (SPEEK) was synthesized by a sulfonation reaction using poly(ether ether ketone) (PEEK). Multilayer structure SPEEK membranes with

  10. Numerical simulations on the rotating flow of wrapped wired HPLWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kissa, A., E-mail: kissa@reak.bme.hu [Budapest Univ. of Tech. and Economics, Inst. of Nuclear Tech. (NTI), Budapest (Hungary); Laurien, E., E-mail: Laurien@ike.uni-stuttgart.de [Univ. of Stuttgart, Inst. for Nuclear Tech. and Energy Systems (IKE), Stuttgart (Germany); Aszodia, A. [Budapest Univ. of Tech. and Economics, Inst. of Nuclear Tech. (NTI), Budapest (Hungary); Zhu, Y. [Univ. of Stuttgart, Inst. for Nuclear Tech. and Energy Systems (IKE), Stuttgart (Germany)

    2011-07-01

    Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or 'sweeping' flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation. (author)

  11. Numerical simulation on a HPLWR fuel assembly flow with one revolution of wrapped wire spacers

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, Attila; Laurien, Eckart [Univ. of Technology and Economics, Budapest (Hungary). Inst. of Nuclear Techniques; Aszodi, Attila; Zhu, Yu [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme

    2010-08-15

    Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods, the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or 'sweeping' flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation. (orig.)

  12. A guidebook for the operation and maintenance of HANARO seismic monitoring analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, Doo Byung; Kim, Hyung Kyoo

    2003-09-01

    Systems and structures related to HANARO safety are classified as seismic category I. Since 1995, the seismic monitoring system has been utilized for monitoring an earthquake at the HANARO site. The existing seismic monitoring system consists of field sensors and monitoring panel. The analog-type monitoring system with magnetic tape recorder is out-of-date model. In addition, the disadvantage of the existing system is that it does not include signal-analyzing equipment. Therefore, we have improved the analog seismic monitoring system into a new digital Seismic Monitoring Analysis System(SMAS) that can offer precise and detail information of the earthquake signals. This newly developed SMAS is operating at the HANARO instrument room to acquire and analyze the signal of an earthquake. This document is a guidebook for the operation and maintenance of the SMAS. The first chapter gives an outline of the SMAS. The second chapter describes functional capability and specification of the hardware. Chapters 3 and 4 describe starting procedure of the SMAS and how to operate the seismic monitoring program, respectively. Chapter 5 illustrates the seismic analysis algorithm used in the SMAS. The way of operating the seismic analysis program is described in chapter 6. Chapter 7 illustrates the calibration procedure for data acquisition module. Chapter 8 describes the symptoms of common malfunctions and its countermeasure suited to the occasions.

  13. Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

    Science.gov (United States)

    Kim, Myong-Seop; Park, Byung-Gun; Kang, Gi-Doo

    2018-01-01

    A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

  14. A Test Device Module of the Step Motor Driver for HANARO CAR Operation

    Energy Technology Data Exchange (ETDEWEB)

    Im, Yun-Taek; Doo, Seung-Gyu; Shin, Jin-Won; Kim, Ki-Hyun; Choi, Young-San; Lee, Jung-Hee; Kim, Hyung-Kyoo; Lee, Choong-Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The brand-new control system is reliable and has advantages compared with the old control system, and the installed system covers all functional operations of old system. Nevertheless, packaged RTP systems do not include a step motor or driver, and it is necessary to develop a proper test device to check the step motor and driver without using the RTP system. In particular, the operation of a CAR (Control Absorber Rod) requires many complicated procedures. Occasionally, it takes significant time to prepare for a field test. In this work, a test device module for a step motor diver is shown to emulate a HANARO CAR operation, and the test device system architecture, operational principle, and experiment results are presented. A commercial 8-bit μ-processor is applied to implement the device. A portable test device for HANARO CAR operation is presented. An 8-bit μ-controller is used to emulate a HANARO CAR operation. The digital interface, as well as the functional operation, of the test device module matches that of the currently used driver. This device can be used to check the functional validity of the step motor and driver.

  15. Load Flow and Short Circuit Analysis of the Class III Power System of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. K.; Jung, H. S

    2005-12-15

    The planning, design, and operation of electric power system require engineering studies to assist in the evaluation of the system performance, reliability, safety and economics. The Class III power of HANARO supplies power for not only HANARO but also RIPF and IMEF. The starting current of most ac motors is five to ten times normal full load current. The loads of the Class III power are connected in consecutive orders at an interval for 10 seconds to avoid excessive voltage drop. This technical report deals with the load flow study and motor starting study for the Class III power of HANARO using ETAP(Electrical Transient Analyzer Program) to verify the capacity of the diesel generator. Short-circuit studies are done to determine the magnitude of the prospective currents flowing throughout the power system at various time intervals after a fault occurs. Short-circuit studies can be performed at the planning stage in order to help finalize the system layout, determine voltage levels, and size cables, transformers, and conductors. From this study, we verify the short circuit current capacity of air circuit breaker(ACB) and automatic transfer switch(ATS) of the Class III power.

  16. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  17. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  18. The development of a new, neutron, time correlated, interrogation method for measurement of 235U content in LWR fuel assemblies

    Science.gov (United States)

    Menlove, H. O.; Menlove, S. H.; Rael, C. D.

    2013-02-01

    This paper presents the first application of a new technique for the measurement of the 235U content in fresh fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by multiplicity counting. The doubles and triples rates are enhanced by having the trigger events from both the 252Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission (ν) of the 252Cf source is 3.76 and the induced fission ν for 235U is 2.44, so the combined ν is ∼5.2 with one neutron removed by the fission reaction. This high effective ν significantly increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252Cf have been minimized by neutron shielding between the source and the detector. This method of active neutron interrogation has been applied to the measurement of fresh pressurized water reactor (PWR) fuel assemblies. The neutron uranium collar (UNCL) that is routinely used for 235U verification in PWR reactor fuel assemblies is used to compare the TCIF method with the typically used AmLi neutron interrogation source. This paper presents both the experimental verification of the TCIF method for a PWR mockup assembly and the MCMPX simulations to optimize the detector geometry.

  19. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  20. Design and Fluid Dynamic Investigations for a High Performance Light Water Reactor Fuel Assembly

    Science.gov (United States)

    Hofmeister, Jan; Laurin, Eckart; Class, Andreas G.

    2005-11-01

    Within the 5th Framework Program of the European Commission a nuclear light water reactor with supercritical steam conditions has been investigated called High Performance Light Water Reactor (HPLWR). This reactor concept is distinct from conventional light water reactor concepts by the fact, that supercritical water is used to achieve higher core outlet temperatures. The reactor operates with a high system pressure, high heat-up of the coolant within the core, and high outlet temperatures of the coolant resulting in a thermal efficiency of up to 44%. We present the design concept proposed by IKET, and a fluid dynamic problem in the foot piece of the fuel assembly, where unacceptable temperature variations must be omitted.

  1. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  2. Postimpact examinations of three DOP 4 iridium shells from simulant fuel sphere assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, E.M.; Hecker, S.S.

    1975-12-01

    Three fuel sphere assemblies, with thoria in doped iridium containment shells, were examined after a simulated earth impact from an aborted orbital mission of a multihundred-watt thermoelectric heat source. The extent of deformation of each unit was measured. Damage to the containment shells was minimal in comparison to that in undoped iridium. Metallographic sections from critical areas indicated that superficial grain boundary cracking in weld zones and microscopic cracking in regions of maximum diameter had occurred in addition to local thinning and coining. The improved properties of the doped iridium are attributed to the retention of a small grain size and to an additional fracture resistance over iridium of a comparable grain size, imparted by either a change in grain boundary chemistry or the flow characteristics of the doped material.

  3. RANS based CFD methodology for a real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae-Ho, E-mail: jhjeong@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of); Song, Min-Seop [Department of Nuclear Engineering, Seoul National University, 559 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Lee, Kwi-Lim [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of)

    2017-03-15

    Highlights: • This paper presents a suitable way for a practical RANS based CFD methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR. • A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI function in a general-purpose commercial CFD code, CFX. • The RANS based CFD methodology is implemented with high resolution scheme and SST turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC and JNC. Furthermore, the RANS based CFD methodology can be successfully extended to the real scale 217-pin wire-wrapped fuel bundles of KAERI PGSFR. • Three-dimensional thermal-hydraulic characteristics have been also investigated briefly. - Abstract: This paper presents a suitable way for a practical RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). The main purpose of the current study is to support license issue for the KAERI PGSFR core safety and to elucidate thermal-hydraulic characteristics in a 217-pin wire-wrapped fuel assembly of KAERI PGSFR. A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI (General Grid Interface) function in a general-purpose commercial CFD code, CFX. The innovative grid generation method with GGI function can achieve to simulate a real wire shape with minimizing cell skewness. The RANS based CFD methodology is implemented with high resolution scheme in convection term and SST (Shear Stress Transport) turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC (Power reactor and Nuclear fuel

  4. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  5. Study of fuel assemblies for the nuclear reactor GFR; Estudio de ensambles de combustible para el reactor nuclear GFR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2008-07-01

    In the present work the criticality calculations for two models of fuel assembly were realized to study the nuclear reactor cooled by gas (Gas Fast Reactor) of IV Generation. Model 1 is an assembly with hexagonal adjustment of fuel rods with reflector in the axial ends higher and lower, the coolant flows between the rods. Model 2 is an hexagonal assembly type block with spheres dispersion and cylindrical channels for where the coolant with reflector in the axial ends also flows. The materials selected for each component of the assemblies, should be resistant to the radiation of fast neutrons and high operation temperatures, for what in both models the following materials were chosen: a mixture of uranium carbide more plutonium for the fuel; a mixture of silicon carbide in different theoretical density percentages for structures and shieldings; helium gas like coolant and a zirconium carbide mixture like reflector, which fulfill the restrictions of being resistant to the high operation temperatures and means of irradiation. General considerations were taken, which are common parameters to both types of assemblies, like size and materials used in the different parts of each model of assembly. The criticality calculations were obtained with the help of the MCNPx code, based on the Monte Carlo method. It was realized a validation of the atomic density data of each component of the assemblies, to have the certainty of the proportionate values that they were introduced of correct way in the code. The results show that model 1 makes better use of the fissile material in a assembly that has the same dimensions externally. That is to say, that from the only considered viewpoint, the neutron one, model 1 is better than model 2. (Author)

  6. Towards neat methanol operation of direct methanol fuel cells: a novel self-assembled proton exchange membrane.

    Science.gov (United States)

    Li, Jing; Cai, Weiwei; Ma, Liying; Zhang, Yunfeng; Chen, Zhangxian; Cheng, Hansong

    2015-04-18

    We report here a novel proton exchange membrane with remarkably high methanol-permeation resistivity and excellent proton conductivity enabled by carefully designed self-assembled ionic conductive channels. A direct methanol fuel cell utilizing the membrane performs well with a 20 M methanol solution, very close to the concentration of neat methanol.

  7. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  8. Numerical analysis on inlet and outlet sections of a test fuel assembly for a Supercritical Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, Attila, E-mail: kissa@reak.bme.hu; Vágó, Tamás; Aszódi, Attila

    2015-12-15

    Graphical abstract: - Highlights: • SCWR-FQT, first facility works with nuclear fuel cooled by SCW, was analysed. • The inlet and outlet section of the test fuel assembly was investigated by CFD. • Two thermohydraulic problems were revealed, described and analysed. • To solve them design changes were proposed and proven by further analysis. - Abstract: The Supercritical Water Reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). One of the major challenges in the development of a SCWR is to develop materials for the fuel and core structures that will be sufficiently corrosion-resistant to withstand supercritical water conditions. Previously, core, reactor and plant design concept of the European High Performance Light Water Reactor (HPLWR) have been worked out in substantial detail. As the next step, it has been proposed to carry out a fuel qualification test of a small scale fuel assembly in a research reactor under typical prototype conditions. Therefore design and licensing of an experimental facility for the fuel qualification test, including the small scale fuel assembly with four fuel rods, the required coolant loop with supercritical water and safety and auxiliary systems, is the scope of the project “Supercritical Water Reactor—Fuel Qualification Test” (SCWR-FQT). This project is a collaborative project co-funded by the European Commission, which takes advantage of a Chinese—European collaboration. As a sub-task of the SCWR-FQT project, the geometry of inlet and outlet sections of the fuel assembly has to be investigated and optimized according to thermohydraulic considerations such as expected stable and uniform inflow pattern and uniform outflow temperature field conditions. To accomplish this task three dimensional CFD analysis has been performed. During the analysis two main problems were identified. On the one hand, generation of a huge eddy was

  9. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  10. In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); De Baere, Paul [European Commission (Luxembourg). Euratom Safeguards

    2015-04-17

    A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

  11. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  12. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  13. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  14. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  15. Review of the transportation of spent fuel assemblies; Bilanz ueber die Transporte abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-11-15

    In August 1999 the transportation of spent fuel assemblies from the Swiss nuclear power plants to the reprocessing facilities was resumed. Until October 2002, 37 transports were carried out without exceeding the legal limits on contamination and dose rate. 19 transports went to La Hague (4 from Beznau, 9 from Goesgen, 4 from Leibstadt and 2 from Muehleberg), 11 to BNFL in Sellafield (2 from Beznau and 9 from Muehleberg) and 7 to the Central Intermediate Storage for Radioactive Wastes (ZWILAG) in Wuerenlingen (4 from Goesgen and 3 from Leibstadt). Additionally, three deliveries with vitrified highly radioactive wastes from the reprocessing facilities were carried out to ZWILAG, also without exceeding any limits. These pleasing results show that the measures required to be taken, as required by the Federal Agency for the Safety of Nuclear Installations (HSK), have contributed to the safe and contamination-free accomplishment of such transportation. Even if the care taken in the carrying out of such transports may not lessen, HSK sees no reason to reduce the present technical and organisational measures. As far as the registration of the individual doses of the participating railway staff is concerned, the results obtained up till now confirm that the radiological exposure is negligible. From the point of view of radiological protection it is not necessary to pursue the registration of the individual radiological exposition. This is why the HSK no longer requires this measure any more. In France, too, the situation concerning contamination during transportation of spent fuel assemblies has been strikingly and lastingly improved. Contamination is only seldom observed and only in connection with a few identified nuclear power plants. In transfers between France and Germany, Belgium as well as the Netherlands, no exceeding of the contamination limits has been noted since resumption of transportation. During the 33 deliveries of empty containers and containers with

  16. Evaluation of assemblies based on carbon materials modified with dendrimers containing platinum nanoparticles for PEM-fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ledesma-Garcia, J.; Barbosa, R.; Chapman, T.W.; Arriaga, L.G.; Godinez, Luis A. [Centro de Investigacion y Desarrollo Tecnologico en Electroquimica, S.C. Parque Tecnologico Queretaro-Sanfandila, 76703 Pedro Escobedo, Qro. (Mexico)

    2009-02-15

    Polyamidoamine (PAMAM) dendrimer-encapsulated Pt nanoparticles (G4OHPt) are synthesized by chemical reduction and characterized by transmission electronic microscopy. An H{sub 2}-O{sub 2} fuel cell has been constructed with porous carbon electrodes modified with the dendrimer nanocomposites. Electrochemical and physical impregnation methods of electrocatalyst immobilization are compared. The modified surfaces are used as electrodes and gas-diffusion layers in the construction of three different membrane-electrode assemblies (MEAs). The MEAs have been tested in a single polymer-electrolyte membrane-fuel cell at 30 C and 20 psig. The fuel cell is, then characterized by electrochemical impedance spectroscopy and cyclic voltammetry, and its performance evaluated in terms of polarization curves and power profiles. The highest fuel cell performance is reached in the MEA constructed by physical impregnation method. The results are compared with a 32 cm{sup 2} prototype cell using commercial electrocatalyst operated at 80 C, obtaining encouraging results. (author)

  17. Membrane-less cloth cathode assembly (CCA) for scalable microbial fuel cells.

    Science.gov (United States)

    Zhuang, Li; Zhou, Shungui; Wang, Yueqiang; Liu, Chengshuai; Geng, Shu

    2009-08-15

    One of the main challenges for scaling up microbial fuel cell (MFC) technologies is developing low-cost cathode architectures that can generate high power output. This study developed a simple method to convert non-conductive material (canvas cloth) into an electrically conductive and catalytically active cloth cathode assembly (CCA) in one step. The membrane-less CCA was simply constructed by coating the cloth with conductive paint (nickel-based or graphite-based) and non-precious metal catalyst (MnO(2)). Under the fed-batch mode, the tubular air-chamber MFCs equipped with Ni-CCA and graphite-CCA generated the maximum power densities of 86.03 and 24.67 mW m(-2) (normalized to the projected cathode surface area), or 9.87 and 2.83 W m(-3) (normalized to the reactor liquid volume), respectively. The higher power output of Ni-CCA-MFC was associated with the lower volume resistivity of Ni-CCA (1.35 x 10(-2)Omega cm) than that of graphite-CCA (225 x 10(-2)Omega cm). At an external resistance of 100 Omega, Ni-CCA-MFC and graphite-CCA-MFC removed approximately 95% COD in brewery wastewater within 13 and 18d, and achieved coulombic efficiencies of 30.2% and 19.5%, respectively. The accumulated net water loss through the cloth by electro-osmotic drag exhibited a linear correlation (R(2)=0.999) with produced coulombs. With a comparable power production, such CCAs only cost less than 5% of the previously reported membrane cathode assembly. The new cathode configuration here is a mechanically durable, economical system for MFC scalability.

  18. Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waata, C.L.

    2006-07-15

    The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

  19. Assembly and Stacking of Flow-through Enzymatic Bioelectrodes for High Power Glucose Fuel Cells.

    Science.gov (United States)

    Abreu, Caroline; Nedellec, Yannig; Gross, Andrew J; Ondel, Olivier; Buret, Francois; Goff, Alan Le; Holzinger, Michael; Cosnier, Serge

    2017-07-19

    Bioelectrocatalytic carbon nanotube based pellets comprising redox enzymes were directly integrated in a newly conceived flow-through fuel cell. Porous electrodes and a separating cellulose membrane were housed in a glucose/oxygen biofuel cell design with inlets and outlets allowing the flow of electrolyte through the entire fuel cell. Different flow setups were tested and the optimized single cell setup, exploiting only 5 mmol L-1 glucose, showed an open circuit voltage (OCV) of 0.663 V and provided 1.03 ± 0.05 mW at 0.34 V. Furthermore, different charge/discharge cycles at 500 Ω and 3 kΩ were applied to optimize long-term stability leading to 3.6 J (1 mW h) of produced electrical energy after 48 h. Under continuous discharge at 6 kΩ, about 0.7 mW h could be produced after a 24 h period. The biofuel cell design further allows a convenient assembly of several glucose biofuel cells in reduced volumes and their connection in parallel or in series. The configuration of two biofuel cells connected in series showed an OCV of 1.35 V and provided 1.82 ± 0.09 mW at 0.675 V, and when connected in parallel, showed an OCV of 0.669 V and provided 1.75 ± 0.09 mW at 0.381 V. The presented design is conceived to stack an unlimited amount of biofuel cells to reach the necessary voltage and power for portable electronic devices without the need for step-up converters or energy managing systems.

  20. Detailed Design of Cooling Water System for Cold Neutron Source in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Soo; Choi, Jung Woon; Kim, Y. K.; Wu, S. I.; Lee, Y. S

    2007-04-15

    To make cold neutron, a cryogenic refrigerator is necessary to transform moderator into cryogenic state so, thermal neutron is changed into cold neutron through heat transfer with moderator. A cryogenic refrigerator mainly consists of two apparatus, a helium compressor and a cold box which needs supply of cooling water. Therefore, cooling water system is essential to operate of cryogenic refrigerator normally. This report is mainly focused on the detailed design of the cooling water system for the HANARO cold neutron source, and describes design requirement, calculation, specification of equipment and water treatment method.

  1. Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

    Science.gov (United States)

    Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

    2011-10-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

  2. Decay characteristics of once-through LWR and LMFBR spent fuels, high-level wastes, and fuel-assembly structural material wastes

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Alexander, C.W.

    1980-11-01

    The decay characteristics of spent fuel, high-level waste, and fuel-assembly structural material (cladding) waste are presented in the form of ORIGEN2 output tables for (1) a pressurized water reactor operating on a once-through cycle with low-enrichment uranium feed, (2) a boiling-water reactor operating on a once-through cycle with low-enrichment uranium feed, and (3) a liquid-metal fast breeder reactor being fueled with depleted uranium enriched with discharged light water reactor plutonium on a once-through basis. The decay characteristics given include the mass (g), radioactivity (Ci), thermal power (W), photon activity (photons/s and MeV/W-s in 18 energy groups), and neutron activity (neutrons/s) from (..cap alpha..,n) and spontaneous fission events. The first three characteristics are given for each element and for the principal nuclide contributors to the activation products, actinides, and fission products. Also included are a summary description of the ORIGEN2 reactor models that form the basis for the calculated results and a physical description of the fuel assemblies for the three reactors.

  3. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  4. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    Science.gov (United States)

    Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  5. Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system

    Science.gov (United States)

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

  6. Performance enhancement of membrane electrode assemblies with plasma etched polymer electrolyte membrane in PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yong-Hun; Yoon, Won-Sub [School of Advanced Materials Engineering, Kookmin University, 861-1 Jeongneung-dong, Seongbuk-gu, Seoul 136-702 (Korea); Bae, Jin Woo; Cho, Yoon-Hwan; Lim, Ju Wan; Ahn, Minjeh; Jho, Jae Young; Sung, Yung-Eun [World Class University (WCU) program of Chemical Convergence for Energy and Environment (C2E2), School of Chemical and Biological Engineering, College of Engineering, Seoul National University (SNU), 599 Gwanak-Ro, Gwanak-gu, Seoul 151-744 (Korea); Kwon, Nak-Hyun [Fuel Cell Vehicle Team 3, Advanced Technology Center, Corporate Research and Development Division, Hyundai-Kia Motors, 104 Mabuk-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-912 (Korea)

    2010-10-15

    In this work, a surface modified Nafion 212 membrane was fabricated by plasma etching in order to enhance the performance of a membrane electrode assembly (MEA) in a polymer electrolyte membrane fuel cell. Single-cell performance of MEA at 0.7 V was increased by about 19% with membrane that was etched for 10 min compared to that with untreated Nafion 212 membrane. The MEA with membrane etched for 20 min exhibited a current density of 1700 mA cm{sup -2} at 0.35 V, which was 8% higher than that of MEA with untreated membrane (1580 mA cm{sup -2}). The performances of MEAs containing etched membranes were affected by complex factors such as the thickness and surface morphology of the membrane related to etching time. The structural changes and electrochemical properties of the MEAs with etched membranes were characterized by field emission scanning electron microscopy, Fourier transform-infrared spectrometry, electrochemical impedance spectroscopy, and cyclic voltammetry. (author)

  7. Modeling of water transport through the membrane electrode assembly for direct methanol fuel cells

    Science.gov (United States)

    Xu, C.; Zhao, T. S.; Yang, W. W.

    In this work, a one-dimensional, isothermal two-phase mass transport model is developed to investigate the water transport through the membrane electrode assembly (MEA) for liquid-feed direct methanol fuel cells (DMFCs). The liquid (methanol-water solution) and gas (carbon dioxide gas, methanol vapor and water vapor) two-phase mass transport in the porous anode and cathode is formulated based on classical multiphase flow theory in porous media. In the anode and cathode catalyst layers, the simultaneous three-phase (liquid and vapor in pores as well as dissolved phase in the electrolyte) water transport is considered and the phase exchange of water is modeled with finite-rate interfacial exchanges between different phases. This model enables quantification of the water flux corresponding to each of the three water transport mechanisms through the membrane for DMFCs, such as diffusion, electro-osmotic drag, and convection. Hence, with this model, the effects of MEA design parameters on water crossover and cell performance under various operating conditions can be numerically investigated.

  8. Improving startup performance with carbon mesh anodes in separator electrode assembly microbial fuel cells

    KAUST Repository

    Zhang, Fang

    2013-04-01

    In a separator electrode assembly microbial fuel cell, oxygen crossover from the cathode inhibits current generation by exoelectrogenic bacteria, resulting in poor reactor startup and performance. To determine the best approach for improving startup performance, the effect of acclimation to a low set potential (-0.2V, versus standard hydrogen electrode) was compared to startup at a higher potential (+0.2V) or no set potential, and inoculation with wastewater or pre-acclimated cultures. Anodes acclimated to -0.2V produced the highest power of 1330±60mWm-2 for these different anode conditions, but unacclimated wastewater inocula produced inconsistent results despite the use of this set potential. By inoculating reactors with transferred cell suspensions, however, startup time was reduced and high power was consistently produced. These results show that pre-acclimation at -0.2V consistently improves power production compared to use of a more positive potential or the lack of a set potential. © 2013 Elsevier Ltd.

  9. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    Science.gov (United States)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  10. Improving startup performance with carbon mesh anodes in separator electrode assembly microbial fuel cells.

    Science.gov (United States)

    Zhang, Fang; Xia, Xue; Luo, Yong; Sun, Dan; Call, Douglas F; Logan, Bruce E

    2013-04-01

    In a separator electrode assembly microbial fuel cell, oxygen crossover from the cathode inhibits current generation by exoelectrogenic bacteria, resulting in poor reactor startup and performance. To determine the best approach for improving startup performance, the effect of acclimation to a low set potential (-0.2V, versus standard hydrogen electrode) was compared to startup at a higher potential (+0.2 V) or no set potential, and inoculation with wastewater or pre-acclimated cultures. Anodes acclimated to -0.2 V produced the highest power of 1330±60 mW m(-2) for these different anode conditions, but unacclimated wastewater inocula produced inconsistent results despite the use of this set potential. By inoculating reactors with transferred cell suspensions, however, startup time was reduced and high power was consistently produced. These results show that pre-acclimation at -0.2 V consistently improves power production compared to use of a more positive potential or the lack of a set potential. Copyright © 2013 Elsevier Ltd. All rights reserved.

  11. Treating refinery wastewaters in microbial fuel cells using separator electrode assembly or spaced electrode configurations

    KAUST Repository

    Zhang, Fang

    2014-01-01

    The effectiveness of refinery wastewater (RW) treatment using air-cathode, microbial fuel cells (MFCs) was examined relative to previous tests based on completely anaerobic microbial electrolysis cells (MECs). MFCs were configured with separator electrode assembly (SEA) or spaced electrode (SPA) configurations to measure power production and relative impacts of oxygen crossover on organics removal. The SEA configuration produced a higher maximum power density (280±6mW/m2; 16.3±0.4W/m3) than the SPA arrangement (255±2mW/m2) due to lower internal resistance. Power production in both configurations was lower than that obtained with the domestic wastewater (positive control) due to less favorable (more positive) anode potentials, indicating poorer biodegradability of the RW. MFCs with RW achieved up to 84% total COD removal, 73% soluble COD removal and 92% HBOD removal. These removals were higher than those previously obtained in mini-MEC tests, as oxygen crossover from the cathode enhanced degradation in MFCs compared to MECs. © 2013 Elsevier Ltd.

  12. Design Optimization of HANARO Irradiation Capsule for Long-Term Irradiation Testing

    Directory of Open Access Journals (Sweden)

    Kee Nam Choo

    2018-01-01

    Full Text Available As HANARO has been recently required to support new R&D relevant to future nuclear systems requiring much higher neutron fluence, two types of bottom rod tip of the capsule were preliminarily prepared. The first one is a conventional design made of STS304 and welded using a tungsten inert gas (TIG welding method. The other is a new design made of STS316L and welded using electron beam (EB welding to strengthen the fatigue property of the rod tip. During the out-pile testing, they failed after 40 and 203 days, respectively. The fracture surfaces were examined using microscopes and the maximal applied stresses were estimated. The combination of these stresses was proved to be sufficient to cause a fatigue failure of the rod tip of the capsule. Based on the failure analysis, an optimized design of the rod tip of the capsule was made for long-term irradiation testing. It was designed to improve the welding and fatigue properties, to decrease the applied stress on the rod tip, and to fundamentally eliminate the effect of residual stress due to welding. The newly designed capsule was safely out-pile-tested up to 450 days and will be utilized for HANARO irradiation testing.

  13. Effects of Nafion impregnation using inkjet printing for membrane electrode assemblies in polymer electrolyte membrane fuel cells

    OpenAIRE

    Wang, Zhuqing; Nagao, Yuki

    2014-01-01

    We present a method of using inkjet printing to deposit Nafion ionomer as the transport media onto catalyst layer made into membrane electrode assemblies (MEAs) for polymer electrolyte fuel cells (PEMFCs). This method provides a more suitable mode of controlling the solution deposition than the existing deposition methods such as spray painting. The cyclic voltammetry results also show that the inkjet printing method has better performance than spray painting by improving catalyst efficiency....

  14. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    OpenAIRE

    Caruso Stefano

    2016-01-01

    The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geolog...

  15. Structural assembly effects of Pt nanoparticle-carbon nanotube-polyaniline nanocomposites on the enhancement of biohydrogen fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Hoa, Le Quynh, E-mail: hoa@p.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Sugano, Yasuhito; Yoshikawa, Hiroyuki; Saito, Masato [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Tamiya, Eiichi, E-mail: tamiya@ap.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

    2011-11-30

    Graphical abstract: - Abstract: In this work, we designed various polyaniline (PANI) nanocomposites with platinum (Pt) nanoparticle-decorated multi-walled carbon nanotubes (MWCNTs), employed them as anodic catalysts, and studied their structural assembly effects with regard to enhancing biohydrogen fuel cell performance. Of two proposed structures, the PANI/Pt/MWCNTs multilayer nanocomposites showed superior electrocatalytic activities in the hydrogen oxidation reaction and in fuel cell power density relative to the Pt/MWCNTs-PANI core-shell design. These enhancements were attributed to the active interface formed between the Pt nanoparticles and polyaniline nanofibers, where the higher electronic and ionic conductivities of the thin PANI nanofiber layers in contact with Pt active sites were better than with the PANI bound Pt/MWCNTs. We also investigated the change in the electronic state of the composites and the charge-transfer rate caused by varying the structural assembly. Finally, the role of each catalyst component was examined to understand its individual effect on fuel cell performance and to understand its structural assembly effect on enhanced power density.

  16. ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

  17. Development of Geometry Optimization Methodology with In-house CFD code, and Challenge in Applying to Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. H.; Lee, K. L. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The wire spacer has important roles to avoid collisions between adjacent rods, to mitigate a vortex induced vibration, and to enhance convective heat transfer by wire spacer induced secondary flow. Many experimental and numerical works has been conducted to understand the thermal-hydraulics of the wire-wrapped fuel bundles. There has been enormous growth in computing capability. Recently, a huge increase of computer power allows to three-dimensional simulation of thermal-hydraulics of wire-wrapped fuel bundles. In this study, the geometry optimization methodology with RANS based in-house CFD (Computational Fluid Dynamics) code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI (General Grid Interface) function is developed for in-house CFD code. Furthermore, three-dimensional flow fields calculated with in-house CFD code are compared with those calculated with general purpose commercial CFD solver, CFX. The geometry optimization methodology with RANS based in-house CFD code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI function is developed for in-house CFD code as same as CFX. Even though both analyses are conducted with same computational meshes, numerical error due to GGI function locally occurred in only CFX solver around rod surface and boundary region between inner fluid region and outer fluid region.

  18. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    Energy Technology Data Exchange (ETDEWEB)

    Holcombe, Scott, E-mail: scott.holcombe@ife.no [Institute for Energy Technology – OECD Halden Reactor Project, Halden (Norway); Andersson, Peter; Svärd, Staffan Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Uppsala (Sweden); Hallstadius, Lars [Westinghouse Electric Sweden AB, Fredholmsgatan 22, 72163 Västerås (Sweden)

    2016-11-21

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of {sup 85}Kr in the gas plenum to {sup 137}Cs in the fuel stack. The rod-wise ratio of {sup 85}Kr/{sup 137}Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO{sub 2}, and four rods were from an assembly with a burnup of 26 MWd/kgUO{sub 2}. At the time of the measurements, the

  19. Graphene-Supported Platinum Catalyst-Based Membrane Electrode Assembly for PEM Fuel Cell

    Science.gov (United States)

    Devrim, Yilser; Albostan, Ayhan

    2016-08-01

    The aim of this study is the preparation and characterization of a graphene-supported platinum (Pt) catalyst for proton exchange membrane fuel cell (PEMFC) applications. The graphene-supported Pt catalysts were prepared by chemical reduction of graphene and chloroplatinic acid (H2PtCl6) in ethylene glycol. X-ray powder diffraction, thermogravimetric analysis (TGA) and scanning electron microscopy have been used to analyze structure and surface morphology of the graphene-supported catalyst. The TGA results showed that the Pt loading of the graphene-supported catalyst was 31%. The proof of the Pt particles on the support surfaces was also verified by energy-dispersive x-ray spectroscopy analysis. The commercial carbon-supported catalyst and prepared Pt/graphene catalysts were used as both anode and cathode electrodes for PEMFC at ambient pressure and 70°C. The maximum power density was obtained for the Pt/graphene-based membrane electrode assembly (MEA) with H2/O2 reactant gases as 0.925 W cm2. The maximum current density of the Pt/graphene-based MEA can reach 1.267 and 0.43 A/cm2 at 0.6 V with H2/O2 and H2/air, respectively. The MEA prepared by the Pt/graphene catalyst shows good stability in long-term PEMFC durability tests. The PEMFC cell voltage was maintained at 0.6 V without apparent voltage drop when operated at 0.43 A/cm2 constant current density and 70°C for 400 h. As a result, PEMFC performance was found to be superlative for the graphene-supported Pt catalyst compared with the Pt/C commercial catalyst. The results indicate the graphene-supported Pt catalyst could be utilized as the electrocatalyst for PEMFC applications.

  20. Electricity producing property and bacterial community structure in microbial fuel cell equipped with membrane electrode assembly.

    Science.gov (United States)

    Rubaba, Owen; Araki, Yoko; Yamamoto, Shuji; Suzuki, Kei; Sakamoto, Hisatoshi; Matsuda, Atsunori; Futamata, Hiroyuki

    2013-07-01

    It is important for practical use of microbial fuel cells (MFCs) to not only develop electrodes and proton exchange membranes but also to understand the bacterial community structure related to electricity generation. Four lactate fed MFCs equipped with different membrane electrode assemblies (MEAs) were constructed with paddy field soil as inoculum. The MEAs significantly affected the electricity-generating properties of the MFCs. MEA-I was made with Nafion 117 solution and the other MEAs were made with different configurations of three kinds of polymers. MFC-I equipped with MEA-I exhibited the highest performance with a stable current density of 55 ± 3 mA m⁻². MFC-III equipped with MEA-III with the highest platinum concentration, exhibited the lowest performance with a stable current density of 1.7 ± 0.1 mA m⁻². SEM observation revealed that there were cracks on MEA-III. These results demonstrated that it is significantly important to prevent oxygen-intrusion for improved MFC performance. By comparing the data of DGGE and phylogenetic analyzes, it was suggested that the dominant bacterial communities of MFC-I were constructed with lactate-fermenters and Fe(III)-reducers, which consisted of bacteria affiliated with the genera of Enterobacter, Dechlorosoma, Pelobacter, Desulfovibrio, Propioniferax, Pelosinus, and Firmicutes. A bacterium sharing 100% similarity to one of the DGGE bands was isolated from MFC-I. The 16S rRNA gene sequence of the isolate shared 98% similarity to gram-positive Propioniferax sp. P7 and it was confirmed that the isolate produced electricity in an MFC. These results suggested that these bacteria are valuable for constructing the electron transfer network in MFC. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  1. Optimizing membrane electrode assembly of direct methanol fuel cells for portable power

    Science.gov (United States)

    Liu, Fuqiang

    Direct methanol fuel cells (DMFCs) for portable power applications require high power density, high-energy conversion efficiency and compactness. These requirements translate to fundamental properties of high methanol oxidation and oxygen reduction kinetics, as well as low methanol and water crossover. In this thesis a novel membrane electrode assembly (MEA) for direct methanol fuel cells has been developed, aiming to improve these fundamental properties. Firstly, methanol oxidation kinetics has been enhanced and methanol crossover has been minimized by proper control of ionomer crystallinity and its swelling in the anode catalyst layer through heat-treatment. Heat-treatment has a major impact on anode characteristics. The short-cured anode has low ionomer crystallinity, and thus swells easily when in contact with methanol solution to create a much denser anode structure, giving rise to higher methanol transport resistance than the long-cured anode. Variations in interfacial properties in the anode catalyst layer (CL) during cell conditioning were also characterized, and enhanced kinetics of methanol oxidation and severe limiting current phenomenon were found to be caused by a combination of interfacial property variations and swelling of ionomer over time. Secondly, much effort has been expended to develop a cathode CL suitable for operation under low air stoichiometry. The effects of fabrication procedure, ionomer content, and porosity distribution on the microstructure and cathode performance under low air stoichiometry are investigated using electrochemical and surface morphology characterizations to reveal the correlation between microstructure and electrochemical behavior. At the same time, computational fluid dynamics (CFD) models of DMFC cathodes have been developed to theoretically interpret the experimental results, to investigate two-phase transport, and to elucidate mechanism of cathode mixed potential due to methanol crossover. Thirdly, a MEA with low

  2. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model and simulation code for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Myung, H. K.; Yang, S. Y.; Kim, B. H.; Song, J. H.; Oh, J. Z. [Kookmin University, Seoul (Korea)

    2002-03-01

    The flow through a nuclear rod bundle with mixing vanes is very complex and so required a suitable turbulence model for its accurate prediction. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to investigate performance of prediction about turbulence model contained in STAR-CD code and to develop suitable turbulence model which can predict complex flow in nuclear assembly. For several nonlinear {kappa}-{epsilon} turbulence models, their performance were investigated in the prediction of the flow in nuclear fuel assembly, and also their problems were discussed in detail. The results obtained from the present research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 19 refs., 32 figs., 3 tabs. (Author)

  3. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-11-21

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

  4. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  5. Development of radioactive stent using HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kyung Bae; Kim, J. R.; Han, H. S.; Shin, B. C.; Kim, Y. M.; Cho, U. K.; Han, K. H.; Park, W. W.; Chung, Y. J

    1997-10-01

    Radioactive cylindrical was prepared by neutron irradiation of pre-made non-radioactive `1`6`5 Ho-sleeve, which was made by casting polyurethane solution containing `1`6`5`Ho(NO{sub 3}){sub 3} in THF+DMF (10:1) solvent in cylindrical glass tube. Its length and diameter could be easily controlled by glass tube used as a mold. The radioactive stent assembly (`1`6`5Ho-SA) was prepared by covering the metallic stent with radioactive sleeve and then treated both ends with epoxy glue for prevention of peeling off the radioactive sleeve from stent (post-irradiation method). Other preparation method of radioactive stent is similar to that of the first one except using radioactive `1`6`6Ho(NO{sub 3}){sub 3} and glass tube fitted with metallic stent before casting (pre-irradiation method). Scanning electron microscopy and autoradiography exhibited that the distribution of `1`6`5Ho and `1`6`6Ho(NO{sub 3}) compounds in polyurethane matrix was nearly homogeneous. The present preparation methods of radioactive sleeve and stent are quite different from conventional method which metallic stent is coated or implanted with radionuclide. The ease with which the radioactive stent can be prepared and its homogeneous radiation emission make it an attractive radiation applicator for the treatment of esophagus cancer. As an animal studies, 6 pathologic specimens were obtained. An animal with 4 mCi of `1`6`6Ho-SA showed loss of epithelial tissue and inflammation at the submucosal layer 4 weeks after the procedure. Considerable improvement of the inflammatory reaction was observed 7 weeks post-therapy without complication. In case treated with 6 mCi of `1`6`6Ho-SA, tissue destruction and widening of the esophageal lumen were observed and the inflammatory reaction propagated into the muscle layer. In case with 9 mCi of `1`6`6Ho-SA, severe esophagitis with cellular proliferation were seen, which resulted in further narrowing of the lumen. (author). 58 refs., 5 tabs., 14 figs

  6. Compound effects of operating parameters on burnup credit criticality analysis in boiling water reactor spent fuel assemblies

    Directory of Open Access Journals (Sweden)

    Shang-Chien Wu

    2018-02-01

    Full Text Available This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor (keff versus burnup (B are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE14 10 × 10 boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC-68 storage cask. The results revealed that the curves of keff versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of keff,Δk in some compound effects was not a summation of the all Δk resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of keff versus B for both single and compound effects.

  7. A Depletion Evaluation of PLUS7 16X16 Nuclear Fuel Assembly using TRITON

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Park, Kwang Heon; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    A nuclear criticality safety evaluation that applies burnup credit (BUC) to a DSC is performed mainly through a two-step process: (1) the determination of isotopic compositions within UNFs to be loaded into a DSC by a depletion analysis and (2) the determination of the k{sub eff} value with respect to the DSC by a criticality analysis . In particular, isotopic compositions by a depletion analysis should be estimated accurately because the nuclides, especially major actinides such as uranium or plutonium, contained in a UNF has a significant influence on a depletion analysis and the subsequent criticality analysis. One way for an accurate estimation of isotopic compositions is to apply accurate cross section libraries to a depletion analysis. In this work, the new one-group cross section libraries of the ORIGEN code were generated with respect to a domestic nuclear fuel assembly (NFA), PLUS7, with and without gadolinia rods using the SCALE 6.1/TRITON code. Then, one of the new libraries was applied to the SCALE 6.1/STARBUCS code to evaluate the effect of the k{sub eff} value. From the results calculated in these conditions, the following conclusions are drawn. (1) The fission and absorption cross-sections of two fissile nuclides generated from burnups of low-enriched NFAs had higher values than those of high-enriched NFAs, while the fission and absorption cross-sections of U-238 generated from burnups of low-enriched NFAs had lower values than those of high-enriched NFAs. (2) The fission and absorption cross-sections below approximately 15,000 MWD/MTU depended on the amount of gadolinia rods and the initial enrichment, while those above approximately 15,000 MWD/MTU depended on the only initial enrichment. (3) The applications of the new PLUS7 16X16 cross section library gave larger k{sub eff} values than the CE 16X16 cross-section library, except for a few cases. The effects of the new library on the k{sub eff} values ranged from -135 to 434 pcm in reactivity.

  8. Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

    Science.gov (United States)

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

    1982-05-13

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  9. High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

    Science.gov (United States)

    Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

    2012-01-01

    Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

  10. Improvement of data processing of the tritium monitor installed at the HANARO stack

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myong Seop; Seol, Chang Woo; Lee, Hyung Sub; Lee, Choong Sung; Park, Sang Jun; Hong, Kwong Pyo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Tritium, particle, iodine and noble gas detectors have been installed at the HANARO stack, and they are continuously being operated to monitor the abnormal emissions of the above radioactivities. The tritium concentration in the air from the reactor hall is the lower level than the reliably detectable activity concentrations by the monitor. Thus the measured tritium concentration has easily been affected by the detector backgrounds and the activity of ambient noble gas. Since it is impossible to increase the sensitivity of the monitor so far, we decided to improve the processing of the output data from the monitor. First of all, we analyzed all data from the monitor. And, we concluded that an improvement would be possible by using the data from the RMS (Regional Monitoring System) noble gas monitor. Thus, in this paper, the methodology for producing the reliable data from the stack tritium monitor is described.

  11. Change of Tensile Properties of High Cr Ferritic/Martensitic Stainless Steel after Irradiation at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Cho, Hae Dong; Han, Chang Hee; Ahn, S. B.; Choo, K. N

    2005-12-15

    In present study, we evaluated the irradiation properties of high Cr ferritic/martensitic steels. These steels were irradiated in HANARO for 14 days at 297{+-}5 .deg. C, and for 15 days at 307{+-}5 .deg. C, and then the fluence was 2.9x10{sup 20}n/cm{sup 2} (E>1.0Mev). High temperature tensile test after irradiation was performed at hot cell in IMEF. Tensile test temperature range was from room temperature to 700 .deg. C. The yield and ultimate tensile strength of specimens increased, and the elongation of specimens decreased by neutron irradiation. Especially elongation was greatly decreased. As the tensile test temperature increased, the increase of strength by irradiation was diminished. But elongation was not recovered at high temperature tensile test. As the irradiation fluence increased, the increase of yield and tensile strength became larger. But the elongation was not influenced by the increase of irradiation fluence.

  12. Upgrade of RMS computers for Y2K problems in RX and related building of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Taek; Kim, J. T.; Ham, C. S.; Kim, C. H.; Lee, Bong Jae; Jae, Yoo Kyung

    2000-08-01

    The Objectives of this Project are as follows : - To resolve the problems of Y2k and operation and maintenance of RMS Computers in RX and related Building of HANARO - To upgrade 486 PC to Pentium II PC - To make Windows NT-Based platform for aspects of user - To make an information structure for radiation using ireless and network devices The Contents of the Project are as follows : - To make Windows NT-Based platform for Radiation Monitoring System - To make Software Platform and Environment for the developing the application program - To design and implement Database Structure - To implement RS232c communication program between local indicators and scanning computers - To implement IEEE 802.3 ethernet communication program between scanning computers and RMTs - To implement user interface for radiation monitoring - To test and inspect Y2k problems.

  13. A study on the direct use of spent PWR fuel in CANDU reactors -Development of DUPIC fuel on manufacturing and quality control technology-

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, Hyun Soo; Lee, Yung Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Oxidation/reduction process was established after analysis of the effect of process parameter on the sintering behavior using SIMFUEL. Process equipment was studied more detail and some of process equipment items were designed and procured. The chemical analysing method of fission products and fissile content in DUPIC fuel was studied and the behavior and the characteristics of fission products in fuel was also done. Requirement for irradiation in HANARO was analysed to prepare performance evaluation. 100 figs, 48 tabs, 170 refs. (Author).

  14. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary

  15. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  16. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications

    Science.gov (United States)

    Feliciano-Ramos, Ileana; Casan~as-Montes, Barbara; García-Maldonado, María M.; Menendez, Christian L.; Mayol, Ana R.; Díaz-Vazquez, Liz M.; Cabrera, Carlos R.

    2015-01-01

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and…

  17. Self-assembled nitrogen-doped fullerenes and their catalysis for fuel cell and rechargeable metal-air battery applications.

    Science.gov (United States)

    Noh, Seung Hyo; Kwon, Choah; Hwang, Jeemin; Ohsaka, Takeo; Kim, Beom-Jun; Kim, Tae-Young; Yoon, Young-Gi; Chen, Zhongwei; Seo, Min Ho; Han, Byungchan

    2017-06-08

    In this study, we report self-assembled nitrogen-doped fullerenes (N-fullerene) as non-precious catalysts, which are active for the oxygen reduction reaction (ORR) and oxygen evolution reaction (OER), and thus applicable for energy conversion and storage devices such as fuel cells and metal-air battery systems. We screen the best N-fullerene catalyst at the nitrogen doping level of 10 at%, not at the previously known doping level of 5 or 20 at% for graphene. We identify that the compressive surface strain induced by doped nitrogen plays a key role in the fine-tuning of catalytic activity.

  18. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  19. Thermal-Hydraulic Effect of Pattern of Wire-wrap Spacer in 19-pin Rod Bundle for SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    As sodium-cooled fast reactor (SFR) has been considered the most promising reactor type for future and prototype gen-IV SFR has been developed actively in Korea, thermal-hydraulic aspects of the SFR fuel assembly have the important role for the reactor safety analysis. In PGSFR fuel assembly, 271 pins of fuel rods are tightly packed in triangular array inside hexagonal duct, and wire is wound helically per each fuel rod with regular pattern to assure the gap between rods and prevent the collision, which is called wire-wrapped spacer. Due to helical shape of the wire-wrapped spacer, flow inside duct can have stronger turbulent characteristics and thermal mixing effect. However, many studies showed the possible wake from swirl flow inside subchannel, which cause local hot spot. To prevent the wake flow and improve thermal mixing, new pattern of wire wrap spacer was suggested. To evaluate the effect of wire wrap spacer pattern, CFD analysis was performed for 19-pin rod bundle with comparison of conventional and U-pattern wire wrap spacer. To prevent the wake due to same direction of swirl flow, 7-rod unit pattern of wire spacer, which are arranged to have different rotational direction of wire with adjacent rods and center rod without wire wrap was proposed. From simulation results, swirl flow across gap conflicts its rotation direction causing wake flow from the regular pattern of the conventional one, which generates local hot spot near cladding. With U-pattern of wire wrap spacer, heat transfer in subchannel can be enhanced with evenly distributed cross flow without compensating pressure loss. From the results, the pattern of wire wrap spacer can influence the both heat transfer characteristics and pressure drop, with flow structures generated by wire wrap spacer.

  20. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  1. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  2. Study on the process system of fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Dae Young; Lee, Chung Young; Sim, Bong Shick; Kim, Jun Yun

    2001-04-01

    The project, 'Feasibility review on the installation of FTL in HANARO' charged by KISTEP has been performed for the technical evaluation of the steady state fuel test loop during 4 months from Dec. 2000 to Mar. 2001. This study describes the preliminary process design of FTL. This report includes the followings : - The construction of FTL - The calculation of the process system - The operation conditions of FTL.

  3. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  4. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  5. Design, assembly and operation of polymer electrolyte membrane fuel cell stacks to 1 kW e capacity

    Science.gov (United States)

    Giddey, S.; Ciacchi, F. T.; Badwal, S. P. S.

    Polymer electrolyte membrane (PEM) fuel cell stacks to 1 kW e capacity, with an active area of 225 cm 2 per cell, have been constructed and operated to investigate the fuel quality issues (one of the major barriers for commercialization of this technology), and start/stop, thermal cycling and load following capabilities. The stacks were assembled and tested in stages of 2-, 4-, 8- and 15-cell configurations. This paper describes the design and assembly of the stacks tested, analysis of the results and problems encountered during operation. Though the 1 kW e stack showed a large variation in the temperature of the interconnect plates due to uneven cooling, the individual cell voltages were found to be within 86 mV (under full load). The average power produced by each cell for the 1 kW e stack operating on air/H 2 was 67.5 W (300 mW cm -2). The stack has undergone more than 40 cold start/shut down thermal cycles in the power output range of 0.6-1 kW e over an accumulated operation of ˜300 h with a small degradation in its performance. The electrical efficiency of the stack varied from 39 to 41%. The recoverable combined heat and power (CHP) efficiency of the stack was 65% without external thermal insulation and 80% with external thermal insulation.

  6. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC; Ensamble combustibles con matrices inertes como recargas del ciclo 11 de la Unidad 1 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO{sub 2} and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO{sub 2}-CeO{sub 2}, 34 fuel bars with UO{sub 2} and 16 fuel bars with UO{sub 2}-Gd{sub 2O}3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO{sub 2} and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO{sub 2} and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  7. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  8. Self assembled 12-tungstophosphoric acid-silica mesoporous nanocomposites as proton exchange membranes for direct alcohol fuel cells.

    Science.gov (United States)

    Tang, Haolin; Pan, Mu; Jiang, San Ping

    2011-05-21

    A highly ordered inorganic electrolyte based on 12-tungstophosphoric acid (H(3)PW(12)O(40), abbreviated as HPW or PWA)-silica mesoporous nanocomposite was synthesized through a facile one-step self-assembly between the positively charged silica precursor and negatively charged PW(12)O(40)(3-) species. The self-assembled HPW-silica nanocomposites were characterized by small-angle XRD, TEM, nitrogen adsorption-desorption isotherms, ion exchange capacity, proton conductivity and solid-state (31)P NMR. The results show that highly ordered and uniform nanoarrays with long-range order are formed when the HPW content in the nanocomposites is equal to or lower than 25 wt%. The mesoporous structures/textures were clearly presented, with nanochannels of 3.2-3.5 nm in diameter. The (31)P NMR results indicates that there are (≡SiOH(2)(+))(H(2)PW(12)O(40)(-)) species in the HPW-silica nanocomposites. A HPW-silica (25/75 w/o) nanocomposite gave an activation energy of 13.0 kJ mol(-1) and proton conductivity of 0.076 S cm(-1) at 100 °C and 100 RH%, and an activation energy of 26.1 kJ mol(-1) and proton conductivity of 0.05 S cm(-1) at 200 °C with no external humidification. A fuel cell based on a 165 μm thick HPW-silica nanocomposite membrane achieved a maximum power output of 128.5 and 112.0 mW cm(-2) for methanol and ethanol fuels, respectively, at 200 °C. The high proton conductivity and good performance demonstrate the excellent water retention capability and great potential of the highly ordered HPW-silica mesoporous nanocomposites as high-temperature proton exchange membranes for direct alcohol fuel cells (DAFCs).

  9. Analysis of core stability measurement data of advanced 9 x 9 fuel assembly in a BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, Katsuhiro; Itami, Akira; Kubo, Yuichiro; Shakudo, Taketomi [Nuclear Fuel Industries Ltd., Tokyo (Japan); Kreuter, D.; Anegawa, Takafumi; Kitamura, Hideya; Ishikawa, Masumi

    1997-05-01

    The core stability measurements were taken during the cycle-9 startup of the 1,300 MWe BWR, Kernkraftwerk Kruemmel (KKK). The core contained advanced 9 x 9 type high burn-up design reload fuel with a higher enrichment than current 8 x 8 fuel. A design feature of the advanced 9 x 9 fuel assembly (FA) is a large square water channel for enhanced neutron moderation. The measurement data as a function of core flow and power showed almost the same stability characteristics as those of the past measurement during the cycle-3 startup of the KKK core with the 8 x 8 FA. The local power range monitors (LPRM) detected neutron flux oscillations in both core-wide in-phase and half-core out-of-phase modes. The frequency-domain stability analysis using the STAIF-PK code well reproduced the measurement result that the onset of unstable operation in KKK first occurs when about half of the reactor internal pumps are operating and the other half are stopped. The stability performance of the advanced 9 x 9 FA in the core was compared with the 8 x 8 FA by a design parameter analysis with respect to thermal-hydraulic and neutronic design. It has been demonstrated by the analysis that the stability performance of the advanced 9 x 9 FA is comparable with current 8 x 8 FA. (author)

  10. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels.

    Science.gov (United States)

    Kucharski, Timothy J; Ferralis, Nicola; Kolpak, Alexie M; Zheng, Jennie O; Nocera, Daniel G; Grossman, Jeffrey C

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  11. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels

    Science.gov (United States)

    Kucharski, Timothy J.; Ferralis, Nicola; Kolpak, Alexie M.; Zheng, Jennie O.; Nocera, Daniel G.; Grossman, Jeffrey C.

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol-1, and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  12. Evaluation of the thermal-mechanic performance of fuel rods MOX in fuel assemblies 10 x 10; Evaluacion del desempeno termo-mecanico barras combustibles MOX en ensambles combustible 10 x 10

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H., E-mail: hector.hernandez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In the Instituto Nacional de Investigaciones Nucleares (Mexico) , we have been working in proposals of fuel assemblies that bear to the reduction of the plutonium inventories that exist a global level, plutonium coming from the dismantlement of the nuclear weapons as of the one used as fuel inside the reactors in operation at the present time. For this reason besides carrying out the evaluation of the neutron performance is necessary to realize the evaluation of the thermal-mechanic behavior of the rods that compose a fuel assembly with the purpose of determining if under the operation conditions to those that are subjected the fuel does not surpass the limit established and this causes a failure in the fuel element. In this sense when carrying out the analysis of an fuel element of mixed oxides in an arrangement 10 x 10 is observed that under the established operation conditions for the proposed cycle values that surpass the limit established for fuel failure are not presented, therefore the proposed assembly can be used as reload element in the nuclear power plant of Laguna Verde. (Author)

  13. Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

    Science.gov (United States)

    Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

    2018-01-01

    In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

  14. Method for removing solid particulate material from within liquid fuel injector assemblies

    Science.gov (United States)

    Simandl, Ronald F.; Brown, John D.; Andriulli, John B.; Strain, Paul D.

    1998-01-01

    A method for removing residual solid particulate material from the interior of liquid fuel injectors and other fluid flow control mechanisms having or being operatively associated with a flow-regulating fixed or variable orifice. The method comprises the sequential and alternate introduction of columns of a non-compressible liquid phase and columns of a compressed gas phase into the body of a fuel injector whereby the expansion of each column of the gas phase across the orifice accelerates the liquid phase in each trailing column of the liquid phase and thereby generates turbulence in each liquid phase for lifting and entraining the solid particulates for the subsequent removal thereof from the body of the fuel injector.

  15. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Joo, W. K.; Kong, D. W.; Park, H. Z. [Yonsei University, Seoul (Korea)

    2001-04-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to develop turbulence model which can predict complex flow. Also, the module will be produced, which can implement the developed turbulence model in the CFX code. The selected turbulence models are k-epsilon model, non-linear k-epsilon model, Reynolds stress model and modified Reynolds stress model to test their performance in the prediction of the flow in nuclear assembly. These models are tested for a 2-D backwise step flow, square duct flow, rod bundle flow and subchannel flow using CFX. The modules, which can implement Reynolds stress model and non-linear k-epsilon odel in CFX code, are produced. The advantages and disadvantages for these turbulence models are described and the limitation of implementation of non-linear model in CFX code is discussed. The results obtained from the research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 18 refs., 37 figs., 8 tabs. (Author)

  16. Development and Application of a Sample Holder for In Situ Gaseous TEM Studies of Membrane Electrode Assemblies for Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Kamino, Takeo; Yaguchi, Toshie; Shimizu, Takahiro

    2017-10-01

    Polymer electrolyte fuel cells hold great potential for stationary and mobile applications due to high power density and low operating temperature. However, the structural changes during electrochemical reactions are not well understood. In this article, we detail the development of the sample holder equipped with gas injectors and electric conductors and its application to a membrane electrode assembly of a polymer electrolyte fuel cell. Hydrogen and oxygen gases were simultaneously sprayed on the surfaces of the anode and cathode catalysts of the membrane electrode assembly sample, respectively, and observation of the structural changes in the catalysts were simultaneously carried out along with measurement of the generated voltages.

  17. Experimental study of local coolant hydrodynamics in TVS-Kvadrat PWR reactor fuel assembly using mixing spacer grids with different types of deflectors

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-12-01

    Full Text Available Results of experimental studies of local hydrodynamic characteristics of coolant flow in fuel assemblies of RWR reactors using different types of mixing spacer grids are presented. Specific features and regularities of coolant flow in fuel pin bundles of TVS-KVADRAT fuel assemblies with different types of mixing spacer grids were revealed in the course of experiments. Analysis of space distribution of projections of absolute flow velocity allowed detailed description of coolant flow beyond the spacer grid with installation of three different types of deflectors. Optimal design of deflector for spacer grid of the TVS-KVADRAT fuel assembly in the standard cell in the area of guiding channels was identified. Results of studies of local hydrodynamics of coolant flow in the TVS-KVADRAT fuel assembly are accepted for subsequent practical application by the JSC Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM in the evaluations of thermal engineering reliability of PWR reactor cores and were included in the database for verification of computational fluid dynamic codes (CFD-codes and implementation of detailed cell array calculations of PWR reactor cores.

  18. Characterizing Flow-Induced Vibrations of Fuel Assemblies for Future Liquid Metal Cooled Nuclear Reactors Using Quasi-Distributed Fibre-Optic Sensors

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2017-08-01

    Full Text Available Excessive vibration of nuclear reactor components, such as the heat exchanger or the fuel assembly should be avoided as these can compromise the lifetime of these components and potentially lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. However, identifying adequate sensors or techniques that can be successfully applied to record the vibrations of the components in a flow of liquid metal at elevated temperatures is very challenging. In this paper, we demonstrate the precise measurements of the vibrations of a very representative mock-up of a fuel assembly in a lead-bismuth eutectic cooled installation using quasi-distributed fibre Bragg grating (FBG based sensors. The unique properties of these sensors, in combination with a dedicated integration and mounting approach, allows for accounting of the severe geometrical constraints and allows characterizing the vibration of the fuel assembly elements under nominal operation conditions. To that aim, we instrumented a single fuel pin within the fuel assembly with 84 FBGs, and conducted spectral measurements with an acquisition rate of up to 5000 measurements per second, enabling the monitoring of local strains of a few με. These measurements provide the information required to assess vibration-related safety hazards.

  19. Structural Integrity Evaluation of an New In-Chimney Bracket Structures for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Jung, Hoan Sung; Seo, Choon Gyo; Shin, Jin Won

    2007-12-15

    In HANARO are there provided three hexagonal irradiation holes (CT, IR1 and IR2) in the central region of the core while four circular irradiation holes (OR3 {approx} OR6) in the outer core. There exist two types of irradiation facilities: uninstrumented or instrumented. The uninstrumented irradiation facility is little influenced by the coolant flow. But the dynamic behavior by the flow-induced vibration (FIV) and seismic loads is expected to largely occur in case of the instrumented test facility due to the long guide tube to protect the instrumentation cables. To suppress this dynamic behavior of the facility, the in-chimney bracket was designed. As a supplementary supporting structure for irradiation facility, this bracket will hold guide tubes whose holding position of the instrumented facility in CT or IR is the middle part of the instrumented facility between the hole spider and the robot arm already provided in the reactor pool liner. On the while, the bracket will grip the upper part of the guide tube when it is applied to hold the instrumented facility loaded in OR sites. Therefore it is believed that the irradiation test can be successfully conducted since this bracket can reduce the FIV and dynamic response to seismic load as well. In new in-chimney bracket, IR1 is reserved for IPS(In-Pile Section) so only CT/IR2 guide tubes are supported by CT/IR clamp units and the shape of In-chimney bracket is redesigned. For evaluating the structural integrity on the new in-chimney bracket and related reactor structures, ANSYS finite element analysis model is developed and the dynamic characteristics are analyzed. The seismic response analyses of new in-chimney bracket and related reactor structures of HANARO under the design earthquake response spectrum loads of OBE(0.1g) and SSE(0.2g) are performed. The response shows that the stress values for main points on the reactor structures and the new in-chimney bracket for seismic loads are within the ASME Code limits

  20. In situ observations of water production and distribution in an operating H2/O2 PEM fuel cell assembly using 1H NMR microscopy.

    Science.gov (United States)

    Feindel, Kirk W; LaRocque, Logan P-A; Starke, Dieter; Bergens, Steven H; Wasylishen, Roderick E

    2004-09-22

    Proton NMR imaging was used to investigate in situ the distribution of water in a polymer electrolyte membrane fuel cell operating on H2 and O2. In a single experiment, water was monitored in the gas flow channels, the membrane electrode assembly, and in the membrane surrounding the catalysts. Radial gradient diffusion removes water from the catalysts into the surrounding membrane. This research demonstrates the strength of 1H NMR microscopy as an aid for designing fuel cells to optimize water management.

  1. Investigation of Ruthenium Dissolution in Advanced Membrane Electrode Assemblies for Direct Methanol Based Fuel Cells Stacks

    Science.gov (United States)

    Valdez, T. I.; Firdosy, S.; Koel, B. E.; Narayanan, S. R.

    2005-01-01

    This viewgraph presentation gives a detailed review of the Direct Methanol Based Fuel Cell (DMFC) stack and investigates the Ruthenium that was found at the exit of the stack. The topics include: 1) Motivation; 2) Pathways for Cell Degradation; 3) Cell Duration Testing; 4) Duration Testing, MEA Analysis; and 5) Stack Degradation Analysis.

  2. Non-intrusive Experimental Study on Nuclear Fuel Assembly Response to Seismic Loads

    Science.gov (United States)

    Weichselbaum, Noah A.

    Experimental measurements of nuclear fuel bundle response to seismic loads have primarily been focused on the response of the structure. Forcing methods have included use of shake tables, however, the majority of work has used hydraulic actuators rigidly connected to a single spacer grid to force the fuel bundle. Structural measurements utilize such instruments as linear variable displacement transducers (LVDT) that are mounted on the structure. From these measurements it has been shown that fuel bundles in prototypical conditions, with an axial flow of 6 m/s, behave markedly different from fuel bundles in still water when there is external forcing on the core from an earthquake. It has also been shown that the structure and fluid are fully coupled. Thus more recently attention has been focused on fluid measurements in the bypass region around fuel bundles with external forcing with laser doppler velocimetry (LDV), which is a point wise fluid velocity measurement technique. This work describes a unique facility that has garnered a large experimental database of fully coupled fluid and structure measurements with time resolved particle image velocimetry (PIV) and digital image correlation (DIC) within a full height 6x6 fuel bundle exposed to seismic forcing on a large 6 degree of freedom shake table. A refractive index matched (RIM) vertical liquid tunnel is mounted on the shake table and houses the fuel bundle which is based on the geometry of a prototypical fuel bundle in a pressurized water reactor (PWR). PIV is obtained with high spatial resolution by rigidly mounting all optical equipment to the test section on the shake table, where the laser light is delivered through high power multi-mode step index fiber optics from a high powered Nd:YLF laser located 10 meters away from the test section. High temporal resolution for the PIV measurements is obtained with state of the art high speed CMOS cameras that record straight to hard drive allowing for increased

  3. The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, P. [VTT Technical Research Centre (Finland)

    2003-07-01

    The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

  4. Fast heating of fuel assembled in a spherical deuterated polystyrene shell target by counter-irradiating tailored laser pulses delivered by a HAMA 1 Hz ICF driver

    Science.gov (United States)

    Mori, Y.; Nishimura, Y.; Hanayama, R.; Nakayama, S.; Ishii, K.; Kitagawa, Y.; Sekine, T.; Takeuchi, Y.; Kurita, T.; Satoh, N.; Kawashima, T.; Komeda, O.; Nishi, T.; Azuma, H.; Hioki, T.; Motohiro, T.; Sunahara, A.; Sentoku, Y.; Miura, E.

    2017-11-01

    Fast heating is a method of heating an assembled high-density plasma into a hot state by irradiating it with short-duration (sub-picosecond), high-intensity (> 1018 W cm-2 ) laser pulses before the plasma expands and dissolves hydrodynamically. In this paper, we present detailed experimental results of fast heating fuel assembled in a spherical deuterated polystyrene shell target of 500 μ m diameter and 7 μm thickness with counterbeam illumination by using a HAMA 1 Hz, 5.9 J inertial confinement fusion laser driver with pulse tailoring. These tailored pulses contain three pulses in sequence: a ‘foot’ pulse of 2.4 J/25 ns, a ‘spike’ pulse of 0.5 J/300 ps and a ‘heater’ pulse of 0.4 J/110 fs; these pulses are designed to assemble the fuel and heat it. By varying the energy of the foot pulse, we find that fast heating the fuel is achieved only if the fuel is weakly ablated by the foot pulse and then shock-assembled by the spike pulse into the target centre so that the heater pulse can access the fuel with a focal intensity greater than 1018 W cm-2 . Without a foot pulse, the heater pulse contributes to assembling the fuel. For higher foot-pulse energies, the heater pulse drives a hydrodynamic motion with speeds of the order 107 cm s-1 with intensities of the order 1017 W cm-2 , resulting in re-assembling and additional heating of the pre-assembled fuel. Once a shock-assembled core is achieved at the target centre, we succeed qualitatively in fast heating the core for shots in sequence with variations of laser energy within 18%. The coupling efficiency from the heating laser to the core is inferred to be (10 +/- 2) % in total: (8 +/- 1.6) % for the ionized bulk electrons and (2 +/- 0.4) % for the bulk ions. The fusion neutron spectrum detected on the laser axis exhibits peaks at 1.0 MeV, 1.7 MeV and 3.8 MeV. These peaks are attributed to the C(d, n){\\hspace{0pt}}13 N and d(d, n){\\hspace{0pt}}3 He reactions induced by counterpropagating fast deuterons

  5. Durability of Membrane Electrode Assemblies (MEAs) in PEM Fuel Cells Operated on Pure Hydrogen and Oxygen

    Science.gov (United States)

    Stanic, Vesna; Braun, James; Hoberecht, Mark

    2003-01-01

    Proton exchange membrane (PEM) fuel cells are energy sources that have the potential to replace alkaline fuel cells for space programs. Broad power ranges, high peak-to-nominal power capabilities, low maintenance costs, and the promise of increased life are the major advantages of PEM technology in comparison to alkaline technology. The probability of PEM fuel cells replacing alkaline fuel cells for space applications will increase if the promise of increased life is verified by achieving a minimum of 10,000 hours of operating life. Durability plays an important role in the process of evaluation and selection of MEAs for Teledyne s Phase I contract with the NASA Glenn Research Center entitled Proton Exchange Membrane Fuel cell (PEMFC) Power Plant Technology Development for 2nd Generation Reusable Launch Vehicles (RLVs). For this contract, MEAs that are typically used for H2/air operation were selected as potential candidates for H2/O2 PEM fuel cells because their catalysts have properties suitable for O2 operation. They were purchased from several well-established MEA manufacturers who are world leaders in the manufacturing of diverse products and have committed extensive resources in an attempt to develop and fully commercialize MEA technology. A total of twelve MEAs used in H2/air operation were initially identified from these manufacturers. Based on the manufacturers specifications, nine of these were selected for evaluation. Since 10,000 hours is almost equivalent to 14 months, it was not possible to perform continuous testing with each MEA selected during Phase I of the contract. Because of the lack of time, a screening test on each MEA was performed for 400 hours under accelerated test conditions. The major criterion for an MEA pass or fail of the screening test was the gas crossover rate. If the gas crossover rate was higher than the membrane intrinsic permeability after 400 hours of testing, it was considered that the MEA had failed the test. Three types of

  6. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  7. Self-assembly of highly charged polyelectrolyte complexes with superior proton conductivity and methanol barrier properties for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Yilmaztuerk, Serpil; Deligoez, Hueseyin; Yilmazoglu, Mesut; Damyan, Hakan; Oeksuezoemer, Faruk; Koc, S. Naci; Durmus, Ali; Ali Guerkaynak, M. [Istanbul University, Engineering Faculty, Chemical Engineering Dept., 34320 Avcilar-Istanbul (Turkey)

    2010-02-01

    The paper is concerned with the formation of Layer-by-Layer (LbL) self-assembly of highly charged polyvinyl sulfate potassium salt (PVS) and polyallylamine hydrochloride (PAH) on Nafion membrane to obtain the multilayered composite membranes with both high proton conductivity and methanol blocking properties. Also, the influences of the salt addition to the polyelectrolyte solutions on membrane selectivity (proton conductivity/methanol permeability) are discussed in terms of controlled layer thickness and charge density. The deposition of the self-assembly of PAH/PVS is confirmed by SEM analysis and it is observed that the polyelectrolyte layers growth on each side of Nafion membrane regularly. (PAH/PVS){sub 10}-Na{sup +} and (PAH/PVS){sub 10}-H{sup +} with 1.0 M NaCl provide 55.1 and 43.0% reduction in lower methanol permittivity in comparison to pristine Nafion, respectively, while the proton conductivities are 12.4 and 78.3 mS cm{sup -1}. Promisingly, it is found that the membrane selectivity values ({phi}) of all multilayered composite membranes in H{sup +} form are much higher than those of Na{sup +} form and perfluorosulfonated ionomers reported in the literature. These encouraging results indicate that composite membranes having both superior proton conductivity and improved methanol barrier properties can be prepared from highly charged polyelectrolytes including salt for fuel cell applications. (author)

  8. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  9. Model-Based Control of a Continuous Coating Line for Proton Exchange Membrane Fuel Cell Electrode Assembly

    Directory of Open Access Journals (Sweden)

    Vikram Devaraj

    2015-01-01

    Full Text Available The most expensive component of a fuel cell is the membrane electrode assembly (MEA, which consists of an ionomer membrane coated with catalyst material. Best-performing MEAs are currently fabricated by depositing and drying liquid catalyst ink on the membrane; however, this process is limited to individual preparation by hand due to the membrane’s rapid water absorption that leads to shape deformation and coating defects. A continuous coating line can reduce the cost and time needed to fabricate the MEA, incentivizing the commercialization and widespread adoption of fuel cells. A pilot-scale membrane coating line was designed for such a task and is described in this paper. Accurate process control is necessary to prevent manufacturing defects from occurring in the coating line. A linear-quadratic-Gaussian (LQG controller was developed based on a physics-based model of the coating process to optimally control the temperature and humidity of the drying zones. The process controller was implemented in the pilot-scale coating line proving effective in preventing defects.

  10. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

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

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  11. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  12. Intermediate review on the transportation of spent fuel assemblies; Zwischenbilanz ueber die Transporte abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-15

    The transportation of spent fuel from the Swiss nuclear power plants to the reprocessing facilities in France and England was interrupted in May 1998 because of contamination that occurred. These measures were presented in the March 1999 statement made by the Office for the Safety of Nuclear Plants (HSK). The transport of spent fuel has been once more permitted and carried out under new conditions since August 1999. In its interim report of October 2000, HSK analyses and evaluates the experience gained since the resumption of transports. For each measure required, it compares the advantages and drawbacks and makes decisions on the maintenance or reduction of the measures to be taken. Between August 1999 and July 2000, 12 spent fuel transports were carried out between the Swiss nuclear power plants and the COGEMA reprocessing facility in France (7 from Goesgen, 4 from Beznau and 1 from Leibstadt). Neither noticeable disagreement with nor exceeding of contamination limits were noted during those 12 transports. This satisfactory result demonstrates that the measures required to be taken are effective. HSK expected from the measures a reduction of the frequency of exceeding contamination limits to less than 5% and also a marked reduction in their frequency. The present results correspond to this expectation; however, the statistical basis is not yet sufficient to be able to draw definitive conclusions. Nevertheless it is noticed that the situation in France, where similar measures have been taken, was very clearly improved. The frequency of exceeding contamination limits was reduced to 2% during the first semester of the year 2000, while it amounted to more than 30% before April 1998. It is the comprehensiveness of the measures required by HSK which allows the avoidance of contamination. The analysis shows that just a small number of measures only contribute insignificantly to the goal sought after. Therefore, two measures will be suppressed (packing of the empty

  13. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    Energy Technology Data Exchange (ETDEWEB)

    Bardet, Philippe [George Washington Univ., Washington, DC (United States); Ricciardi, Guillaume [Atomic Energy Commission (CEA) (France)

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  14. Environmentally Friendly Recycling of Fuel-Cell Membrane Electrode Assemblies by Using Ionic Liquids.

    Science.gov (United States)

    Balva, Maxime; Legeai, Sophie; Leclerc, Nathalie; Billy, Emmanuel; Meux, Eric

    2017-07-21

    The platinum nanoparticles used as the catalyst in proton exchange membrane fuel cells (PEMFCs) represent approximately 46 % of the total price of the cells for a large-scale production, and this is one of the barriers to their commercialization. Therefore, the recycling of the platinum catalyst could be the best alternative to limit the production costs of PEMFCs. The usual recovery routes for spent catalysts containing platinum are pyro-hydrometallurgical processes in which a calcination step is followed by aqua regia treatment, and these processes generate fumes and NOx emissions, respectively. The electrochemical recovery route proposed here is more environmentally friendly, performed under "soft" temperature conditions, and does not result in any gas emissions. It consists of the coupling of the electrochemical leaching of platinum in chloride-based ionic liquids (ILs), followed by its electrodeposition. The leaching of platinum was studied in pure ILs and in ionic-liquid melts at different temperatures and with different chloride contents. Through the modulation of the composition of the ionic-liquid melts, it is possible to leach and electrodeposit the platinum from fuel-cell electrodes in a single-cell process under an inert or ambient atmosphere. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  15. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  16. Transient signal generation in a self-assembled nanosystem fueled by ATP

    Science.gov (United States)

    Pezzato, Cristian; Prins, Leonard J.

    2015-07-01

    A fundamental difference exists in the way signal generation is dealt with in natural and synthetic systems. While nature uses the transient activation of signalling pathways to regulate all cellular functions, chemists rely on sensory devices that convert the presence of an analyte into a steady output signal. The development of chemical systems that bear a closer analogy to living ones (that is, require energy for functioning, are transient in nature and operate out-of-equilibrium) requires a paradigm shift in the design of such systems. Here we report a straightforward strategy that enables transient signal generation in a self-assembled system and show that it can be used to mimic key features of natural signalling pathways, which are control over the output signal intensity and decay rate, the concentration-dependent activation of different signalling pathways and the transient downregulation of catalytic activity. Overall, the reported methodology provides temporal control over supramolecular processes.

  17. Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Goodsell, Alison Victoria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-14

    Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDA instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.

  18. Electrode assembly for use in a solid polymer electrolyte fuel cell

    Science.gov (United States)

    Raistrick, Ian D.

    1989-01-01

    A gas reaction fuel cell may be provided with a solid polymer electrolyte membrane. Porous gas diffusion electrodes are formed of carbon particles supporting a catalyst which is effective to enhance the gas reactions. The carbon particles define interstitial spaces exposing the catalyst on a large surface area of the carbon particles. A proton conducting material, such as a perfluorocarbon copolymer or ruthenium dioxide contacts the surface areas of the carbon particles adjacent the interstitial spaces. The proton conducting material enables protons produced by the gas reactions adjacent the supported catalyst to have a conductive path with the electrolyte membrane. The carbon particles provide a conductive path for electrons. A suitable electrode may be formed by dispersing a solution containing a proton conducting material over the surface of the electrode in a manner effective to coat carbon surfaces adjacent the interstitial spaces without impeding gas flow into the interstitial spaces.

  19. Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

    Energy Technology Data Exchange (ETDEWEB)

    Na, Chongzheng [Univ. of Notre Dame, IN (United States)

    2016-10-17

    Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety of synthetic methods, we show that synthesizing graphene-­supported UO2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.

  20. An Artificial Gravity Spacecraft Approach which Minimizes Mass, Fuel and Orbital Assembly Reg

    Science.gov (United States)

    Bell, L.

    2002-01-01

    The Sasakawa International Center for Space Architecture (SICSA) is undertaking a multi-year research and design study that is exploring near and long-term commercial space development opportunities. Space tourism in low-Earth orbit (LEO), and possibly beyond LEO, comprises one business element of this plan. Supported by a financial gift from the owner of a national U.S. hotel chain, SICSA has examined opportunities, requirements and facility concepts to accommodate up to 100 private citizens and crewmembers in LEO, as well as on lunar/planetary rendezvous voyages. SICSA's artificial gravity Science Excursion Vehicle ("AGSEV") design which is featured in this presentation was conceived as an option for consideration to enable round-trip travel to Moon and Mars orbits and back from LEO. During the course of its development, the AGSEV would also serve other important purposes. An early assembly stage would provide an orbital science and technology testbed for artificial gravity demonstration experiments. An ultimate mature stage application would carry crews of up to 12 people on Mars rendezvous missions, consuming approximately the same propellant mass required for lunar excursions. Since artificial gravity spacecraft that rotate to create centripetal accelerations must have long spin radii to limit adverse effects of Coriolis forces upon inhabitants, SICSA's AGSEV design embodies a unique tethered body concept which is highly efficient in terms of structural mass and on-orbit assembly requirements. The design also incorporates "inflatable" as well as "hard" habitat modules to optimize internal volume/mass relationships. Other important considerations and features include: maximizing safety through element and system redundancy; means to avoid destabilizing mass imbalances throughout all construction and operational stages; optimizing ease of on-orbit servicing between missions; and maximizing comfort and performance through careful attention to human needs. A

  1. Prediction of critical heat flux in fuel assemblies using a CHF table method

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; Hwang, Dae Hyun; Bang, Je Geon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Baek, Won Pil; Chang, Soon Heung [Korea Advance Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor. 8 refs., 3 figs., 3 tabs. (Author)

  2. Neutron Shielding Calculation of a Beam Stopper for the Thermal-TAS at HANARO using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byoungil; Kim, Jongsoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    It can be classified as a virtual source, neutron optics components, Monochromator Shielding Unit (MSU), Sample table and Analyzer and Detector bank. At a monochromator, specific wavelength of neutron beam is selected by the Bragg scattering theory. Because the neutron beam comes from reactor source has high intensity, massive shielding units are placed around monochromator. After monochromator, selected neutron beam goes toward sample table. This is a brief account of the Thermal-TAS experiment. To achieve high quality experimental results, signal noise should be controlled. Signal noise can come from not only electronic hardware system but also un-necessary neutron background. To reduce background, proper shielding components should be placed around optical components. Monochromator, analyzer and detector have their own shielding components but sample table doesn't in current status of Thermal-TAS. To reduce background occurred on sample table, direct neutron beam that comes from monochromator should be blocked. In this paper, basic design and radiation shielding calculation of direct beam stopper shielding component have been conducted. A conceptual design and shielding calculation of a neutron beam stopper for the thermal-TAS at HANARO has been conducted. With this result, a beam stopper segment will be fabricated. After further procedures such as mechanical, electronic design and fabrication of a driving part of the stopper, a neutron beam stopper system will be installed on the thermal-TAS at HANARO.

  3. Development and demonstration of an advanced extended-burnup fuel-assembly design incorporating urania-gadolinia. Second semi-annual progress report, October 1981-March 1982

    Energy Technology Data Exchange (ETDEWEB)

    Newman, L W; Rombough, C T; Thornton, T A

    1982-08-01

    The Babcock and Wilcox Company, Duke Power Company, and the US Department of Energy are participating in an extended-burnup program for pressurized water reactors that will demonstrate an advanced fuel assembly design. This advanced fuel assembly will use a UO/sub 2/-Gd/sub 2/O/sub 3/ burnable-poison fuel mixture along with other state-of-the-art fuel performance and uranium utilization-enhancing design features that include annular pellets, annealed guide tubes, Zircaloy intermediate grids, and removable upper end fittings. Comparisons of the thermal properties of UO/sub 2/-Gd/sub 2/O/sub 3/ specimens containing 2.98, 5.66, and 8.50 wt % Gd/sub 2/O/sub 3/ with UO/sub 2/ specimens showed that thermal conductivity is the only thermal parameter significantly affected by the addition of Gd/sub 2/O/sub 3/. The milling steps used to prepare UO/sub 2/-Gd/sub 2/O/sub 3/ powder result in a powder that is more active than standard UO/sub 2/ powder. As a result, UO/sub 2/-Gd/sub 2/O/sub 3/ fuel has shown more variability than UO/sub 2/ fuel in as-sintered theoretical density and densification behavior. However, a poreforming material, added to the UO/sub 2/-Gd/sub 2/O/sub 3/ powder mixture before sintering, can be used to achieve the desired density. Measured results from critical experiments were compared with predicted data and confirmed the accuracy of the standard two-group diffusion theory model for predicting global and discrete UO/sub 2/-Gd/sub 2/O/sub 3/ effects when cross-section input is appropriately adjusted. The preliminary first two fuel cycles for lead test assemblies of the advanced design were developed. Irradiation of the lead test assemblies is scheduled to begin in 1983 in Duke Power Company's Oconee Unit 1. An intercalibrated movable incore detector system will be used to monitor the performance of the test assemblies during irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-01

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

  5. A joint study on the research reactor development technology in Brazil, Vietnam, Thailand, etc. and a study on the establishment of basic foundation for the activation of the HANARO technology export

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C. W.; Lee, J. B.; Kim, H. R. [and others

    2003-03-01

    According to the characteristics of this project, the results of the technical cooperation development are the agreed terms, developed items, and others that the one who is conducting a research reactor project, requires from Hanaro specifically.

  6. Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Batou, A., E-mail: anas.batou@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Soize, C., E-mail: christian.soize@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Brie, N., E-mail: nicolas.brie@edf.fr [EDF R and D, Département AMA, 1 avenue du général De Gaulle, 92140 Clamart (France)

    2013-09-15

    Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading.

  7. Evaluation of the potential of various aquatic eco-systems in harnessing bioelectricity through benthic fuel cell: effect of electrode assembly and water characteristics.

    Science.gov (United States)

    Venkata Mohan, S; Srikanth, S; Veer Raghuvulu, S; Mohanakrishna, G; Kiran Kumar, A; Sarma, P N

    2009-04-01

    Six different types of ecological water bodies were evaluated to assess their potential to generate bioelectricity using benthic type fuel cell assemblies. Experiments were designed with various combinations of electrode assemblies, surface area of anode and anodic materials. Among the 32 experiments conducted, nine combinations evidenced stable electron-discharge/current. Nature, flow conditions and characteristics of water bodies showed significant influence on the power generation apart from electrode assemblies, surface area of anode and anodic material. Stagnant water bodies showed comparatively higher power output than the running water bodies. Placement of cathode on algal mat (as bio-cathode) documented several folds increment in power output. Electron-discharge started at 1000 Omega resistance in polluted water bodies (Nacaharam cheruvu, Hussain Sagar lake Musi river), whereas, in relatively less polluted water bodies (Uppal pond/stream, Godavari river) electron-discharge was observed at low resistances (500/750 Omega).

  8. Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

    2018-01-01

    This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

  9. The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Xi [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Xiao, Zejun, E-mail: fabulous_2012@sina.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Yan, Xiao; Li, Yongliang; Huang, Yanping [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China)

    2013-05-15

    Highlights: ► CFX and MCNP codes are suitable to calculate the axial power profile of the FA. ► The partition method in the calculation will affect the final result. ► The density feedback has little effect on the axial power profile of CSR1000 FA. -- Abstract: SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.

  10. Structured multilayered electrodes of proton/electron conducting polymer for polymer electrolyte membrane fuel cells assembled by spray coating

    Energy Technology Data Exchange (ETDEWEB)

    Wolz, Andre; Zils, Susanne; Roth, Christina [Institute for Materials Science, TU Darmstadt, Petersenstr. 23, D-64287 Darmstadt (Germany); Michel, Marc [Department of Advanced Materials and Structures, CRP Henri Tudor, 66 Rue de Luxembourg, L-4002 Esch-sur-Alzette (Luxembourg)

    2010-12-15

    Membrane electrode assemblies (MEAs) for fuel cell applications consist of electron conductive support materials, proton conductive ionomer, and precious metal nanoparticles to enhance the catalytic activity towards H{sub 2} oxidation and O{sub 2} reduction. An optimized connection of all three phases is required to obtain a high noble metal utilization, and accordingly a good performance. Using polyaniline (PANI) as an alternative support material, the generally used ionomer Nafion {sup registered} could be replaced in the catalyst layer. PANI has the advantage to be electron and proton conductive at the same time, and can be used as a catalyst support as well. In this study, a new technique building up alternating layers of PANI supported catalyst and single-walled carbon nanotubes (SWCNT) supported catalyst is introduced. Multilayers of PANI and SWCNT catalysts are used on the cathode side, whereas the anode side is composed of commercial platinum/carbon black catalyst and Nafion {sup registered}, applied by an airbrush. No additional Nafion {sup registered} ionomer is used for proton conductivity of the cathode. The so called spray coating method results in high power densities up to 160 mW cm{sup -2} with a Pt loading of 0.06 mg cm{sup -2} at the cathode, yielding a Pt utilization of 2663 mW mg{sub Pt}{sup -1}. As well as PANI, supports of SWCNTs have the advantage to have a fibrous structure and additional, they provide high electron conductivity. The combination of the new technique and the fibrous 1-dimensional support materials leads to a porous 3-dimensional electrode network which could enhance the gas transport through the electrode as well as the Pt utilization. The spray coating method could be upgraded to an in-line process and is not restricted to batch production. (author)

  11. CFD ANALYSIS OF THE SPACER GRIDS AND MIXING VANES EFFECT ON THE FLOW IN THE CHOSEN PART OF THE TVSA-T FUEL ASSEMBLY

    Directory of Open Access Journals (Sweden)

    Jakub Juklíček

    2015-10-01

    Full Text Available CFD is a promising and widely spread tool for a flow simulation in nuclear reactor fuel assemblies. One of the limiting factors is the complicated geometry of a spacer grid. It leads to the computational mesh with high number of cells and with possibility of decreasing quality. Therefore an approach to simulate the flow as precisely as possible and simultaneously in a reasonable computational expense has to be chosen. The goal of the following CFD analysis is to obtain the detailed velocity field in a precise geometry of a chosen part of the TVSA-T fuel assembly. This kind of simulation provides data for comparison that can be applied in many situations, for instance, for comparison with simulations when a porous media boundary condition is applied as a replacement of the spacer grid.TVSA-T fuel assembly is equipped with combined spacer grids. Combined spacer grid has two functions - support of the fuel pins as a part of assembly skeleton and mixing vanes which ensures coolant mixing. The support part is geometrically very complicated and it is impossible to prepare a good quality computational mesh there. It is also difficult to create a mesh in the support part and the mixing part joint area because of inaccurate connection between these two parts.A representative part of the TVSA-T fuel assembly with a combined spacer grid segment was chosen to perform the CFD simulation. Some inevitable geometry simplifications of the spacer grid geometry were performed. These simplifications were as insignificant as possible to preserve the flow character and to make it possible to prepare a quality mesh at the same time.Steady state CFD simulation was performed with k-ε realizable turbulence model. Heat transfer was not simulated and only velocity field was investigated. Detailed flow characterization which was obtained from this calculation shown, that mixing vanes already affect the flow in the support part of the grid thanks to suction effect. Vortex

  12. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  13. Development of cancer therapy facility of HANARO and medical research in BNCT; development of the technique for boron concentration analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hee Dong; Byun, Soo Hyun; Sun, Gwang Min; Kim, Suk Kwon; Kim, In Jung; Park, Chang Su [Seoul National University, Seoul (Korea)

    2002-03-01

    Objective and Necessity of the Project- Development of a boron concentration analysis facility used for BNCT. - Development of the technique for boron concentration analysis. Contents and Scopes of the Project - Construction of the boron concentration analysis facility based on PGAA. Estimation of the neutron beam characteristics. -Establishment of the technique for the boron concentration analysis. - Estimation of the reliability for the boron analysis. Results of the Project -Installation of the boron concentration analysis facility at Hanaro. - Neutron beam characteristics are the sample position (neutron flux : 7.9 x 10{sup 7} n/cm{sup 2}s, Cd-ratio : 266) Technique for the boron concentration analysis. - Boron detection sensitivity and limit (detection sensitivity : 2, 131 cps/mg-B, detection limit : 67 ng for 10,000 sec). 63 refs., 37 figs., 13 tabs. (Author)

  14. Study on the characterization of the neutron radiography facility in HANARO for two-phase flow research

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Seo, C. G.; Jeong, J. H.; Lee, B. H.; Choi, Y. S

    2001-01-01

    For the application of dynamic neutron radiography to the two-phase flow research using HANARO, several experimental items to which the radiography technique is beneficial were identified through the review of the outputs from the related researches and the discussions with experts. Also, the investigation of the equipments including the beam port, camera and converter was made and a hardware and a software for image processing were equipped. It was confirmed that the calibration curve for the attenuation of neutron beam in fluid which is required for the two-phase flow experiment could be obtained by the computer code calculation. Based on the investigation results on the equipment and the results from the measurement of BNCT beam characteristics, a high speed camera and an image intensifier will be purchased. Then, the high speed dynamic neutron radiography facility for two-phase flow experiments will be fully equipped.

  15. Characterization of HANARO neutron radiography facility in accordance with ASTM standard E545-91/E803-91 for KOLAS/ISO17025.

    Science.gov (United States)

    Cheul-Muu, Sim; Ki-Yong, Nam; In-Cheol, Lim; Chang-Hee, Lee; Ha-Lim, Choi

    2004-10-01

    As neutron radiography is even more in demand for industrial applications of aircraft, turbine blade, automobile, explosive igniters, etc, it is necessary to review the standards which are the most appropriate for preparing the procedures for setting up the QA system. Recently, Korea Of Lab Accreditation Scheme (KOLAS) was originated from ISO 17025. It is widely recognized by research peer groups for conducting valid tests. The neutron radiography facility (NRF) of High Flux Advanced Neutron Application Reactor (HANARO), which started ion 1996, is the preliminary stages of KOLAS. The HANARO NRF is not only characterized using ASTM standards E545-91/E803-91 to satisfy the requirements of KOLAS, but in the design phase of the tomography system.

  16. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  17. Design, Fabrication and Test Report on HANARO Instrumented Capsule (08M-01K) for the Evaluation of Irradiation Degradation of RPV Model Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Cho, M. S.; Kim, B. G.; Kang, Y. H.; Son, J. M.; Shin, Y. T.; Park, S. J.; Oh, S. Y.; Lee, J. H

    2009-03-15

    An instrumented capsule of 08M-01K was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of RPV model alloys. The basic structure of the 08M-01K capsule was based on the 07M-13N capsule which was successfully irradiated in the CT test hole of HANARO. 124 specimens such as PCVN, 1/2 PCVN, Charpy, small tensile, PA and TEM specimens of SA508 RPV model alloys were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe-Ag-Nb-Co neutron fluence monitors installed in the capsule. A friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 255{approx} 317 .deg. C up to a fast neutron fluence of 8.3x10{sup 19} n/cm{sup 2} (E>1.0 MeV). The obtained results will be very valuable for the evaluation on the extended operation of the KORI 1 Nuclear Power Reactor.

  18. Design, Fabrication and Test Report on Instrumented Capsule (08M-02K) for Irradiation Test of RPV Model Alloys in HANARO OR5 Test Hole

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Cho, M. S.; Kim, B. G.; Kang, Y. H.; Son, J. M.; Shin, Y. T.; Park, S. J.; Oh, S. Y

    2009-09-15

    An instrumented capsule of 08M-02K was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of RPV model alloys. The basic structure of the 08M-02K capsule was based on the 07M-21K capsule which was successfully irradiated in the OR5 test hole of HANARO. 228 specimens such as PCVN, 1/2 PCVN, Charpy, small tensile and TEM specimens of SA508 RPV model alloys were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. A friction welded tube between STS304 and AI1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during capsule cutting process in HANARO. The capsule was irradiated in the OR5 test hole of HANARO of a 30MW thermal output at 269{approx}310 .deg. C up to a fast neutron fluence of 4.4x10{sup 19}(n/cm{sup 2}) (E>1.0 MeV). The obtained results will be very valuable for the evaluation on the 2nd extended operation of the KORI 1 Nuclear Power Reactor.

  19. A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2010-12-06

    There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

  20. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  1. Conjugate heat transfer study of a wire spacer SFR fuel assembly thanks to the thermal code SYRTHES and the CFD code Code_Saturne

    Science.gov (United States)

    Péniguel, C.; Rupp, I.; Rolfo, S.; Hermouet, D.

    2014-06-01

    The paper presents a HPC calculation of a conjugate heat transfer simulation in fuel assembly as those found in liquid metal coolant fast reactors. The wire spacers, helically wound along each pin axis, generate a strong secondary flow pattern in opposition to smooth pins. Assemblies with a range of pins going from 7 to 271 have been simulated, 271 pins corresponding to the industrial case. Both the fluid domain, as well as the solid part, are detailed leading to large meshes. The fluid is handled by the CFD code Code_Saturne using 98 million cells, while the solid domain is taken care of thanks to the thermal code SYRTHES on meshes up to 240 million cells. Both codes are fully parallel and run on cluster with hundreds of processors. Simulations allow access to the temperature field in nominal conditions and degraded situations.

  2. ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

    2007-10-01

    Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of

  3. Advanced Manufacturing of Intermediate Temperature, Direct Methane Oxidation Membrane Electrode Assemblies for Durable Solid Oxide Fuel Cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed innovation builds on the successes of the Phase I program by integrating our direct oxidation membrane electrode assembly (MEA) into a monolithic solid...

  4. Air Breathing Direct Methanol Fuel Cell

    Science.gov (United States)

    Ren; Xiaoming

    2003-07-22

    A method for activating a membrane electrode assembly for a direct methanol fuel cell is disclosed. The method comprises operating the fuel cell with humidified hydrogen as the fuel followed by running the fuel cell with methanol as the fuel.

  5. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  6. Direct Measurement of Initial Enrichment and Burn-up of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    Energy Technology Data Exchange (ETDEWEB)

    Henzl, Vladimir [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

    2012-07-16

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to utilize non-destructive assay (NDA) techniques to determine the elemental plutonium (Pu) content in a commercial-grade nuclear spent fuel assembly (SFA). In the third year of the NGSI Spent Fuel NDA project, the research focus is on the integration of a few NDA techniques. One of the reoccurring challenges to the accurate determination of Pu content has been the explicit dependence of the measured signal on the presence of neutron absorbers which build up in the assembly in accordance with its operating and irradiation history. The history of any SFA is often summarized by the parameters of burn-up (BU), initial enrichment (IE) and cooling time (CT). While such parameters can typically be provided by the operator, the ability to directly measure and verify them would significantly enhance the autonomy of the IAEA inspectorate. Within this paper, we demonstrate that an instrument based on a Differential Die-Away technique is in principle capable of direct measurement of IE and, should the CT be known, also the BU.

  7. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR; Estudio del comportamiento neutronico de un ensamble combustible con gadolinia de un reactor HPLWR

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    This work presents a neutronic study of a square assembly design of double line of fuel rods, with moderator box to center of the arrangement, for the nuclear reactor cooled with supercritical water, High Performance Light Water Reactor (HPLWR). For the fuel analyses of the reactor HPLWR the neutronic code Helios-2 was used, settling down as the first study on fuel under conditions of supercritical water that has been simulated with this code. The analyzed variables, essentials in the neutronic design of any reactor, were the infinite neutrons multiplication factor (k{infinity}) and the maximum power peaking factor (PPF{sub max}), as well as the reactivity coefficients by the fuel temperature. The k{infinity} and PPF{sub max} values were obtained under conditions in cold (293.6 K) and in hot (to 880.8 K). The tests were realized for a reference fuel assembly design, with 40 fuel rods with enrichments of 4 and 5% of U-235, and considering different concentrations of consumable poison (gadolinium - Gd{sub 2O3}) in some rods of the same assembly. The obtained results show values k{infinity} and PPF{sub max} minors to the present in the conventional light water reactors. Moreover, the reactivity coefficients by fuel temperature were verified with the purpose of satisfying the safety conditions required in the nuclear reactors. (Author)

  8. Controlling Heat Release from a Close-Packed Bisazobenzene-Reduced-Graphene-Oxide Assembly Film for High-Energy Solid-State Photothermal Fuels.

    Science.gov (United States)

    Zhao, Xiaoze; Feng, Yiyu; Qin, Chengqun; Yang, Weixiang; Si, Qianyu; Feng, Wei

    2017-04-10

    A closed-cycle system for light-harvesting, storage, and heat release is important for utilizing and managing renewable energy. However, combining a high-energy, stable photochromic material with a controllable trigger for solid-state heat release remains a great challenge for developing photothermal fuels (PTFs). This paper presents a uniform PTF film fabricated by the assembly of close-packed bisazobenzene (bisAzo) grafted onto reduced graphene oxide (rGO). The assembled rGO-bisAzo template exhibited a high energy density of 131 Wh kg-1 and a long half-life of 37 days owing to inter- or intramolecular H-bonding and steric hindrance. The rGO-bisAzo PTF film released and accumulated heat to realize a maximum temperature difference (DT) of 15 °C and a DT of over 10 °C for 30 min when the temperature difference of the environment was greater than100 °C. Controlling heat release in the solid-state assembly paves the way to develop highly efficient and high-energy PTFs for a multitude of applications. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D. F.; Rochman, D.; Koning, A. J. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {sup 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)

  10. Analyzing Structural Changes of Fe-N-C Cathode Catalysts in PEM Fuel Cell by Mößbauer Spectroscopy of Complete Membrane Electrode Assemblies.

    Science.gov (United States)

    Kramm, Ulrike I; Lefèvre, Michel; Bogdanoff, Peter; Schmeißer, Dieter; Dodelet, Jean-Pol

    2014-11-06

    The applicability of analyzing by Mößbauer spectroscopy the structural changes of Fe-N-C catalysts that have been tested at the cathode of membrane electrode assemblies in proton exchange membrane (PEM) fuel cells is demonstrated. The Mößbauer characterization of powders of the same catalysts was recently described in our previous publication. A possible change of the iron species upon testing in fuel cell was investigated here by Mößbauer spectroscopy, energy-dispersive X-ray cross-sectional imaging, and neutron activation analysis. Our results show that the absorption probability of γ rays by the iron nuclei in Fe-N-C is strongly affected by the presence of Nafion and water content. A detailed investigation of the effect of an oxidizing treatment (1.2 V) of the non-noble cathode in PEM fuel cell indicates that the observed activity decay is mainly attributable to carbon oxidation causing a leaching of active iron sites hosted in the carbon matrix.

  11. The influence of membrane electrode assembly water content on the performance of a polymer electrolyte membrane fuel cell as investigated by 1H NMR microscopy.

    Science.gov (United States)

    Feindel, Kirk W; Bergens, Steven H; Wasylishen, Roderick E

    2007-04-21

    The relation between the performance of a self-humidifying H(2)/O(2) polymer electrolyte membrane fuel cell and the amount and distribution of water as observed using (1)H NMR microscopy was investigated. The integrated (1)H NMR image signal intensity (proportional to water content) from the region of the polymer electrolyte membrane between the catalyst layers was found to correlate well with the power output of the fuel cell. Several examples are provided which demonstrate the sensitivity of the (1)H NMR image intensity to the operating conditions of the fuel cell. Changes in the O(2)(g) flow rate cause predictable trends in both the power density and the image intensity. Higher power densities, achieved by decreasing the resistance of the external circuit, were found to increase the water in the PEM. An observed plateau of both the power density and the integrated (1)H NMR image signal intensity from the membrane electrode assembly and subsequent decline of the power density is postulated to result from the accumulation of H(2)O(l) in the gas diffusion layer and cathode flow field. The potential of using (1)H NMR microscopy to obtain the absolute water content of the polymer electrolyte membrane is discussed and several recommendations for future research are provided.

  12. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  13. Study of gas phase in prototype fuel assemblies with compact uranium dioxide, irradiated in SM-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kirillovich, A.P.; Sulaberidze, V.Sh.; Pimonov, Yu.I.; Shulimov, V.N.; Lavrinovich, Yu.G.; Biryukov, A.S.; Kupriyanov, V.N.

    1984-08-01

    A comprehensive procedure was developed for studying the gas liberation from nuclear fuel. In fuel elements with compact uranium dioxide, Ar, H/sub 2/, O/sub 2/, CO, CO/sub 2/ and N/sub 2/ are present in addition to the gaseous fission products and the introduced helium; following irradiation, the quantity of them is several times greater than initially. The Xe and Kr yield under the jacket of fuel elements is 15-30 percent, while the measured Xe/Kr ratio is 20 percent greater than the calculated value because of the reaction /sup 135/Xe-/sup 136/Xe. The data obtained may be used for physical and process design calculations, as well as for more precisely specifying the procedures for determining gas liberation.

  14. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  15. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  16. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  17. Neutron beam applications - Development of single crystal structure analysis technique using the HANARO neutron four circle diffractometer

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Il Hwan; Kim, Moon Jib; Kim, Jin Gyu [Chungnam National University, Taejon (Korea)

    2000-04-01

    As the four circle diffractometer (FCD) has been set up in HANARO, it has become possible to study the single crystal structures by means of the neutron diffraction. Taking account of the geometry of the FCD, a program for the control of te FCD and neutron data acquisition operating under Windows' circumstance has been accomplished. Also, a computer program which can automatically measure the diffraction intensity data has been developed. All data obtained from the FCD are processed automatically for further work and a software for the single crystal structure analyses has been prepared. A KC1 single crystal was selected as first test sample for a structure analysis had been successfully performed on the FCD using in-house developed program and accordingly their functionings with precision were confirmed. For regular single crystal diffraction experiments, the structure analyses of chrysoberyl and Zr(Y)0{sub 1.87} single crystals were performed using both neutron and X-ray diffraction methods, and the result showed that the neutron diffraction work is superior to the X-ray one from the viewpoint of certain crystallographic information obtainable only from the former one. 24 refs., 15 figs., 15 tabs. (Author)

  18. Green-fuel-mediated synthesis of self-assembled NiO nano-sticks for dual applications—photocatalytic activity on Rose Bengal dye and antimicrobial action on bacterial strains

    Science.gov (United States)

    Iyyappa Rajan, P.; Vijaya, J. Judith; Jesudoss, S. K.; Kaviyarasu, K.; Kennedy, L. John; Jothiramalingam, R.; Al-Lohedan, Hamad A.; Vaali-Mohammed, Mansoor-Ali

    2017-08-01

    With aim of promoting the employability of green fuels in the synthesis of nano-scaled materials with new kinds of morphologies for multiple applications, successful synthesis of self-assembled NiO nano-sticks was achieved through a 100% green-fuel-mediated hot-plate combustion reaction. The synthesized NiO nano-sticks show excellent photocatalytic activity on Rose Bengal dye and superior antibacterial potential towards both Gram-positive and Gram-negative bacteria.

  19. Optimum concentration gradient of the electrocatalyst, Nafion® and poly(tetrafluoroethylene) in a membrane-electrode-assembly for enhanced performance of direct methanol fuel cells.

    Science.gov (United States)

    Liu, Jing Hua; Jeon, Min Ku; Lee, Ki Rak; Woo, Seong Ihl

    2010-12-14

    A combinatorial library of membrane-electrode-assemblies (MEAs) which consisted of 27 different compositions was fabricated to optimize the multilayer structure of direct methanol fuel cells. Each spot consisted of three layers of ink and a gradient was generated by employing different concentrations of the three components (Pt catalyst, Nafion® and polytetrafluoroethylene (PTFE)) of each layer. For quick evaluation of the library, a high-throughput optical screening technique was employed for methanol electro-oxidation reaction (MOR) activity. The screening results revealed that gradient layers could lead to higher MOR activity than uniform layers. It was found that the MOR activity was higher when the concentrations of Pt catalyst and Nafion ionomer decreased downward from the top layer to the bottom layer. On the other hand, higher MOR activity was observed when PTFE concentration increased downward from the top to the bottom layer.

  20. High-performance membrane-electrode assembly with an optimal polytetrafluoroethylene content for high-temperature polymer electrolyte membrane fuel cells

    DEFF Research Database (Denmark)

    Jeong, Gisu; Kim, MinJoong; Han, Junyoung

    2016-01-01

    Although high-temperature polymer electrolyte membrane fuel cells (HT-PEMFCs) have a high carbon monoxide tolerance and allow for efficient water management, their practical applications are limited due to their lower performance than conventional low-temperature PEMFCs. Herein, we present a high...... in the electrodes and result in low performance. MEAs with PTFE content of 20 wt% have an optimal pore structure for the efficient formation of electrolyte/catalyst interfaces and gas channels, which leads to high cell performance of approximately 0.5 A cm-2 at 0.6 V.......-performance membrane-electrode assembly (MEA) with an optimal polytetrafluoroethylene (PTFE) content for HT-PEMFCs. Low or excess PTFE content in the electrode leads to an inefficient electrolyte distribution or severe catalyst agglomeration, respectively, which hinder the formation of triple phase boundaries...

  1. Membrane electrode assembly with enhanced platinum utilization for high temperature proton exchange membrane fuel cell prepared by catalyst coating membrane method

    Science.gov (United States)

    Liang, Huagen; Su, Huaneng; Pollet, Bruno G.; Linkov, Vladimir; Pasupathi, Sivakumar

    2014-11-01

    In this work, membrane electrode assemblies (MEAs) prepared by catalyst coating membrane (CCM) method are investigated for reduced platinum (Pt) loading and improved Pt utilization of high temperature proton exchange membrane fuel cell (PEMFC) based on phosphoric acid (PA)-doped poly(2,5-benzimidazole) (AB-PBI) membrane. The results show that CCM method exhibits significantly higher cell performance and Pt-specific power density than that of MEAs prepared with conventional gas diffusion electrode (GDE) under a low Pt loading level. In-suit cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) show that the MEAs prepared by the CCM method have a higher electrochemical surface area (ECSA), low cell ohmic resistance and low charge transfer resistance as compared to those prepared with GDEs at the same Pt loading.

  2. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

  3. Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2011-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  4. Self-assembly of mixed Pt and Au nanoparticles on PDDA-functionalized graphene as effective electrocatalysts for formic acid oxidation of fuel cells.

    Science.gov (United States)

    Wang, Shuangyin; Wang, Xin; Jiang, San Ping

    2011-04-21

    Pt and Au nanoparticles with controlled Pt : Au molar ratios and PtAu nanoparticle loadings were successfully self-assembled onto poly(diallyldimethylammonium chloride) (PDDA)-functionalized graphene (PDDA-G) as highly effective electrocatalysts for formic acid oxidation in direct formic acid fuel cells (DFAFCs). The simultaneously assembled Pt and Au nanoparticles on PDDA-G showed superb electrocatalytic activity for HCOOH oxidation, and the current density associated with the preferred dehydrogenation pathway for the direct formation of CO(2) through HCOOH oxidation on a Pt(1)Au(8)/PDDA-G (i.e., a Pt : Au ratio of 1 : 8) is 32 times higher than on monometallic Pt/PDDA-G. The main function of the Au in the mixed Pt and Au nanoparticles on PDDA-G is to facilitate the first electron transfer from HCOOH to HCOO(ads) and the effective spillover of HCOO(ads) from Au to Pt nanoparticles, where HCOO(ads) is further oxidized to CO(2). The Pt : Au molar ratio and PtAu nanoparticle loading on PDDA-G supports are the two critical factors to achieve excellent electrocatalytic activity of PtAu/PDDA-G catalysts for the HCOOH oxidation reactions.

  5. High-performance membrane electrode assembly with multi-functional Pt/SnO2eSiO2/C catalyst for proton exchange membrane fuel cell operated under low-humidity conditions

    CSIR Research Space (South Africa)

    Hou, S

    2016-06-01

    Full Text Available A novel self-humidifying membrane electrode assembly (MEA) with homemade multifunctional Pt/SnO(sub2)-SiO(sub2)/C as the anode was developed to improve the performance of a proton exchange membrane fuel cell under low humidity. The MEAs' performance...

  6. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Krementz, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rose, David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dunsmuir, Mike [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-02-06

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  7. Flexseal Insulator Test Assembly

    Science.gov (United States)

    Buchanan, Eric

    1995-01-01

    Small-scale version of solid-fuel rocket motor flexseal nozzle bearing assembly instrumented and tested in compression-testing fixture simulating conditions during rocket motor operation described in report.

  8. Design of an Advanced Membrane Electrode Assembly Employing a Double-Layered Cathode for a PEM Fuel Cell.

    Science.gov (United States)

    Kim, GyeongHee; Eom, KwangSup; Kim, MinJoong; Yoo, Sung Jong; Jang, Jong Hyun; Kim, Hyoung-Juhn; Cho, EunAe

    2015-12-23

    The membrane electrolyte assembly (MEA) designed in this study utilizes a double-layered cathode: an inner catalyst layer prepared by a conventional decal transfer method and an outer catalyst layer directly coated on a gas diffusion layer. The double-layered structure was used to improve the interfacial contact between the catalyst layer and membrane, to increase catalyst utilization and to modify the removal of product water from the cathode. Based on a series of MEAs with double-layered cathodes with an overall Pt loading fixed at 0.4 mg cm(-2) and different ratios of inner-to-outer Pt loading, the MEA with an inner layer of 0.3 mg Pt cm(-2) and an outer layer of 0.1 mg Pt cm(-2) exhibited the best performance. This performance was better than that of the conventional single-layered electrode by 13.5% at a current density of 1.4 A cm(-2).

  9. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  10. Development of membrane electrode assembly for high temperature proton exchange membrane fuel cell by catalyst coating membrane method

    Science.gov (United States)

    Liang, Huagen; Su, Huaneng; Pollet, Bruno G.; Pasupathi, Sivakumar

    2015-08-01

    Membrane electrode assembly (MEA), which contains cathode and anode catalytic layer, gas diffusion layers (GDL) and electrolyte membrane, is the key unit of a PEMFC. An attempt to develop MEA for ABPBI membrane based high temperature (HT) PEMFC is conducted in this work by catalyst coating membrane (CCM) method. The structure and performance of the MEA are examined by scanning electron microscopy (SEM), electrochemical impedance spectroscopy (EIS) and I-V curve. Effects of the CCM preparation method, Pt loading and binder type are investigated for the optimization of the single cell performance. Under 160 °C and atmospheric pressure, the peak power density of the MEA, with Pt loading of 0.5 mg cm-2 and 0.3 mg cm-2 for the cathode and the anode, can reach 277 mW cm-2, while a current density of 620 A cm-2 is delivered at the working voltage of 0.4 V. The MEA prepared by CCM method shows good stability operating in a short term durability test: the cell voltage maintained at ∼0.45 V without obvious drop when operated at a constant current density of 300 mA cm-2 and 160 °C under ambient pressure for 140 h.

  11. Physico-chemical study of the degradation of membrane-electrode assemblies in a proton exchange membrane fuel cell stack

    Science.gov (United States)

    Ferreira-Aparicio, P.; Gallardo-López, B.; Chaparro, A. M.; Daza, L.

    A proton exchange membrane fuel cell stack integrated by 8-elements has been evaluated in an accelerated stress test. The application of techniques such as TEM analyses of ultramicrotome-sliced sections of some samples and XRD, XPS and TGA of spent electrodes reveal the effects of several degradation processes contributing to reduce the cells performance. The reduction of the Pt surface area at the cathode is favored by the oxidation of carbon black agglomerates in the catalytic layer, the agglomeration of Pt particles and by the partial dissolution of Pt, which migrates towards the anode and precipitates within the membrane. In the light of the TEM, EDAX and XPS results, two combined effects are probably responsible of the increase of the internal resistance of the stack cells: (i) a lower proton conductivity of the membranes due to the high affinity of the sulfonic acid groups for ions originated from Pt crystallites and other peripherical elements such as the silicone elastomeric gaskets and (ii) the increment of electrically isolated islands in the cathode gas diffusion electrodes resulting from carbon corrosion and the degradation of the perfluorinated polymers. Water accumulation and inhomogeneous gas distribution throughout the stack cells originate different degradation rates in them.

  12. Performance of two different types of anodes in membrane electrode assembly microbial fuel cells for power generation from domestic wastewater

    KAUST Repository

    Hays, Sarah

    2011-10-01

    Graphite fiber brush electrodes provide high surface areas for exoelectrogenic bacteria in microbial fuel cells (MFCs), but the cylindrical brush format limits more compact reactor designs. To enable MFC designs with closer electrode spacing, brush anodes were pressed up against a separator (placed between the electrodes) to reduce the volume occupied by the brush. Higher maximum voltages were produced using domestic wastewater (COD = 390 ± 89 mg L-1) with brush anodes (360 ± 63 mV, 1000 Ω) than woven carbon mesh anodes (200 ± 81 mV) with one or two separators. Maximum power densities were similar for brush anode reactors with one or two separators after 30 days (220 ± 1.2 and 240 ± 22 mW m-2), but with one separator the brush anode MFC power decreased to 130 ± 55 mW m-2 after 114 days. Power densities in MFCs with mesh anodes were very low (<45 mW m-2). Brush anodes MFCs had higher COD removals (80 ± 3%) than carbon mesh MFCs (58 ± 7%), but similar Coulombic efficiencies (8.6 ± 2.9% brush; 7.8 ± 7.1% mesh). These results show that compact (hemispherical) brush anodes can produce higher power and more effective domestic wastewater treatment than flat mesh anodes in MFCs. © 2011 Elsevier B.V. All rights reserved.

  13. Swelling-resistant nuclear fuel

    Science.gov (United States)

    Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  14. Comparison of electrochemical and microbiological characterization of microbial fuel cells equipped with SPEEK and Nafion membrane electrode assemblies.

    Science.gov (United States)

    Suzuki, Kei; Owen, Rubaba; Mok, Joann; Mochihara, Hiroki; Hosokawa, Takuya; Kubota, Hiroko; Sakamoto, Hisatoshi; Matsuda, Atsunori; Tashiro, Yosuke; Futamata, Hiroyuki

    2016-09-01

    Microbial fuel cells equipped with SPEEK-MEA (SPEEK-MFC) and Nafion-MEA (Nafion-MFC) were constructed with organic waste as electron donor and lake sediment as inoculum and were then evaluated comprehensively by electrochemical and microbial analyses. The proton conductivity of SPEEK was several hundreds-fold lower than that of Nafion 117, whereas the oxygen mass and diffusion transfer coefficients of SPEEK were 10-fold lower than those of Nafion 117. It was difficult to predict which was better membrane for MFC based on the feature of membrane. Analyses of polarization curves indicated that the potential of electricity production was similar in both MFCs, as the SPEEK-MFC produced 50-80% of the practical current density generated by the Nafion-MFC. Chronopotentiometry analyses indicated that the Nafion-MEA kept the performance longer than the SPEEK-MEA for long period, whereas performance of both anodes improved on time. Multidimensional scaling analyses based on DGGE profiles revealed the anolytic and biofilm communities of the SPEEK-MFC had developed differently from those of the Nafion-MFC. Clone library analyses indicated that Geobacter spp. represented 6.3% of the biofilm bacterial community in the Nafion-MFC but not detected in the SPEEK-MFC. Interestingly, the clone closely related to Acetobacterium malicum strain HAAP-1, belonging to the homoacetogens, became dominant in both anolytic and biofilm communities of the SPEEK-MFC. It was suggested that the lower proton conductivity of SPEEK-MEA allowed the bacteria closely related to strain HAAP-1 to be dominant specifically in SPEEK-MFC. These results indicated that Nafion-MFC ranked with SPEEK-MFC and that MEAs had strong selective pressure for electricity-producing bacterial community. Copyright © 2016 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  15. Desenvolvimento de conjunto membrana-eletrodos para célula a combustível de metanol direto passiva Development of membrane electrode assembly for passive direct methanol fuel cell

    Directory of Open Access Journals (Sweden)

    Eli Carlos Lisboa Ferreira

    2010-01-01

    Full Text Available Direct methanol fuel cells (DMFCs without external pumps or other ancillary devices for fuel and oxidant supply are known as passive DMFCs and are potential candidates to replace lithium-ion batteries in powering portable electronic devices. This paper presents the results obtained from a membrane electrode assembly (MEA specifically designed for passive DMFCs. Appropriated electrocatalysts were prepared and the effect of their loadings was investigated. Two types of gas diffusion layers (GDL were also tested. The influence of the methanol concentration was analyzed in each case. The best MEA performance presented a maximum power density of 11.94 mW cm-2.

  16. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  17. Self-assembled platinum nanoparticles on sulfonic acid-grafted graphene as effective electrocatalysts for methanol oxidation in direct methanol fuel cells.

    Science.gov (United States)

    Lu, Jinlin; Li, Yanhong; Li, Shengli; Jiang, San Ping

    2016-02-15

    In this article, sulfonic acid-grafted reduced graphene oxide (S-rGO) were synthesized using a one-pot method under mild conditions, and used as Pt catalyst supports to prepare Pt/S-rGO electrocatalysts through a self-assembly route. The structure, morphologies and physicochemical properties of S-rGO were examined in detail by techniques such as atomic force microscope (AFM), transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). The S-rGO nanosheets show excellent solubility and stability in water and the average particle size of Pt nanoparticles supported on S-rGO is ~3.8 nm with symmetrical and uniform distribution. The electrocatalytic properties of Pt/S-rGO were investigated for methanol oxidation reaction (MOR) in direct methanol fuel cells (DMFCs). In comparison to Pt supported on high surface area Vulcan XC-72 carbon (Pt/VC) and Pt/rGO, the Pt/S-rGO electrocatalyst exhibits a much higher electrocatalytic activity, faster reaction kinetics and a better stability. The results indicate that Pt/S-rGO is a promising and effective electrocatalyst for MOR of DMFCs.

  18. Enhanced power production of a membrane electrode assembly microbial fuel cell (MFC) using a cost effective poly [2,5-benzimidazole] (ABPBI) impregnated non-woven fabric filter.

    Science.gov (United States)

    Choi, Soojung; Kim, Jung Rae; Cha, Jaehwan; Kim, Yejin; Premier, Giuliano C; Kim, Changwon

    2013-01-01

    A membrane electrode assembly (MEA) microbial fuel cell (MFC) with a non-woven paper fabric filter (NWF) was investigated as an alternative to a proton exchange membrane (PEM) separator. The MFC with a NWF generated a cell voltage of 545 mV and a maximum power density of 1027 mW/m(3), which was comparable to that obtained from MFCs with a PEM (551 mV, 609 mW/m(3)). The MFC with a NWF showed stable cell performance (550 mV) over 300 days, whereas, the MFC with PEM performance decreased significantly from 551 mV to 415 mV due to biofilm formation and chemical precipitation on the membrane surface. Poly [2,5-benzimidazole] (ABPBI) was evaluated with respect to its capacity to increased proton conductivity and contact between separator and electrodes. The overall performance of the MFC with ABPBI was improved by enhancing the ion conductivity and steric contact, producing 766 mW/m(3) at optimum loading of 50 mg ABPBI/cm(2). Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. Solid phase microbial fuel cell (SMFC) for harnessing bioelectricity from composite food waste fermentation: influence of electrode assembly and buffering capacity.

    Science.gov (United States)

    Mohan, S Venkata; Chandrasekhar, K

    2011-07-01

    Solid phase microbial fuel cells (SMFC; graphite electrodes; open-air cathode) were designed to evaluate the potential of bioelectricity production by stabilizing composite canteen based food waste. The performance was evaluated with three variable electrode-membrane assemblies. Experimental data depicted feasibility of bioelectricity generation from solid state fermentation of food waste. Distance between the electrodes and presence of proton exchange membrane (PEM) showed significant influence on the power yields. SMFC-B (anode placed 5 cm from cathode-PEM) depicted good power output (463 mV; 170.81 mW/m(2)) followed by SMFC-C (anode placed 5 cm from cathode; without PEM; 398 mV; 53.41 mW/m(2)). SMFC-A (PEM sandwiched between electrodes) recorded lowest performance (258 mV; 41.8 mW/m(2)). Sodium carbonate amendment documented marked improvement in power yields due to improvement in the system buffering capacity. SMFCs operation also documented good substrate degradation (COD, 76%) along with bio-ethanol production. The operation of SMFC mimicked solid-sate fermentation which might lead to sustainable solid waste management practices. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. A fully spray-coated fuel cell membrane electrode assembly using Aquivion ionomer with a graphene oxide/cerium oxide interlayer

    Science.gov (United States)

    Breitwieser, Matthias; Bayer, Thomas; Büchler, Andreas; Zengerle, Roland; Lyth, Stephen M.; Thiele, Simon

    2017-05-01

    A novel multilayer membrane electrode assembly (MEA) for polymer electrolyte membrane fuel cells (PEMFCs) is fabricated in this work, within a single spray-coating device. For the first time, direct membrane deposition is used to fabricate a PEMFC by spraying the short-side-chain ionomer Aquivion directly onto the gas diffusion electrodes. The fully sprayed MEA, with an Aquivion membrane 10 μm in thickness, achieved a high power density of 1.6 W/cm2 for H2/air operation at 300 kPaabs. This is one of the highest reported values for thin composite membranes operated in H2/air atmosphere. By the means of confocal laser scanning microscopy, individual carbon fibers from the gas diffusion layer are identified to penetrate through the micro porous layer (MPL), likely causing a low electrical cell resistance in the range of 150 Ω cm2 through the thin sprayed membranes. By spraying a 200 nm graphene oxide/cerium oxide (GO/CeO2) interlayer between two layers of Aquivion ionomer, the impact of the electrical short is eliminated and the hydrogen crossover current density is reduced to about 1 mA/cm2. The peak power density of the interlayer-containing MEA drops only by 10% compared to a pure Aquivion membrane of similar thickness.

  1. The through-plane thermal conductivity and the contact resistance of the components of the membrane electrode assembly and gas diffusion layer in proton exchange membrane fuel cells

    Science.gov (United States)

    Alhazmi, N.; Ingham, D. B.; Ismail, M. S.; Hughes, K.; Ma, L.; Pourkashanian, M.

    2014-12-01

    The thermal conductivity of the components of the membrane electrode assembly (MEA) and GDL must be accurately estimated in order to better understand the heat transfer processes in the proton exchange membrane (PEM) fuel cells. In this study, an experimental investigation has been performed to measure the through-plane thermal conductivity and the contact resistance for a number of gas diffusion layer (GDL) materials. The sensitivity of these quantities to the temperature, PTFE content and micro porous layer (MPL) coating has been undertaken. In addition, the through-plane thermal conductivity of the membrane has been measured and reported as a function of temperature and water content. Further, the through-plane thermal conductivity of the catalyst layer has been determined as a function of temperature and platinum loading. It has been found that the through-plane thermal conductivity of the components of the MEA decreases when the temperature increases, and the through-plane thermal conductivity of the GDL is significantly lower than its in-plane thermal conductivity.

  2. Heat transfer enhancement for spent nuclear fuel assembly disposal packages using metallic void fillers: A prevention technique for solidification shrinkage-induced interfacial gaps

    Science.gov (United States)

    Park, Yongsoo; McKrell, Thomas J.; Driscoll, Michael J.

    2017-06-01

    This study considers replacing the externally accessible void spaces inside a disposal package containing a spent nuclear fuel assembly (SNFA) with high heat conducting metal to increase the effective thermal conductivity of the package and simplify the heat transfer mechanism inside the package by reducing it to a conduction dominant problem. The focus of the study is on preventing the gaps adjacent to the walls of the package components, produced by solidification shrinkage of poured liquid metal. We approached the problem by providing a temporary coating layer on the components to avoid direct build-up of thick metal oxides on their surface to promote metallic bonding at the interfaces under a non-inert environment. Laboratory scale experiments without SNFA were performed with Zn coated low carbon steel canisters and Zamak-3 void filler under two different filling temperature conditions - below and above the melting point of Zn (designated BMP and AMP respectively). Gap formation was successfully prevented in both cases while we confirmed an open gap in a control experiment, which used an uncoated canister. Minor growth of Al-Fe intermetallic phases was observed at the canister/filler interface of the sample produced under the BMP condition while their growth was significant and showed irregularly distributed morphology in the sample produced under the AMP condition, which has a potential to mitigate excessive residual stresses caused by shrinkage prevention. A procedure for the full-scale application was specified based on the results.

  3. Optimization of the performance of polymer electrolyte fuel cell membrane-electrode assemblies: Roles of curing parameters on the catalyst and ionomer structures and morphology

    Energy Technology Data Exchange (ETDEWEB)

    Bose, Anima B.; Shaik, Ruhulla [Department of Mechanical Engineering, Northern Illinois University, DeKalb, IL 60115 (United States); Mawdsley, Jennifer [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2008-07-15

    In order to understand the origin of performance variations in polymer electrolyte membrane fuel cells (PEMFCs), a series of membrane-electrode assemblies (MEAs) with identical electrode layer compositions were prepared using different electrode curing conditions, their performances were evaluated, and their morphologies determined by scanning electron microscopy (SEM). The polarization curves varied markedly primarily due to differences in morphologies of electrodes, which were dictated by the curing processes. The highest performing MEAs (1.46 W cm{sup -2} peak power density at 3.2 A cm{sup -2} and 80 C) were prepared using a slow curing process at a lower temperature, whereas those MEAs prepared using a faster curing process performed poorly (0.1948 W cm{sup -2} peak power density at 440 mA cm{sup -2} and 80 C). The slowly cured MEAs showed uniform electrode catalyst and ionomer distributions, as revealed in SEM images and elemental maps. The relatively faster cured materials exhibited uneven distribution of ionomer with significant catalyst clustering. Collectively, these results indicate that to achieve optimal performance, factors that affect the dynamics of the curing process, such as rate of solvent evaporation, must be carefully controlled to avoid solvent trapping, minimize catalyst coagulation, and promote even distribution of ionomer. (author)

  4. Critical Configuration and Physics Measurements for Beryllium Reflected Assemblies of U(93.15)O₂ Fuel Rods (1.506-cm Pitch and 7-Tube Clusters)

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Murphy, Michael F. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  5. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2014-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  6. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation; Une approche milieu poreux pour la modeisation de l'interaction fluide-structure des assemblages combustibles dans un coeur de reacteur a eau pressurisee: simulation et experimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ricciardi, G.

    2008-10-15

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  7. Impact of neutron thermal scattering laws on the burn-up analysis of supercritical LWR's fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Conti, Andrea

    2011-10-15

    . The third, most naive option is called the ''free gas approximation''. It is the goal of this work to make an estimate of the criticality calculations' inaccuracy due to the inadequate employed physical model and to determine which one of the available models can be the best replacement. The accuracy of criticality calculations referring to the HPLWR is a problem that had already been raised by Waata in 2006. In her Ph.D. thesis Waata reports having carried out MCNP runs referring to an HPLWR fuel element employing the free gas approximation. In her thesis Waata explicitly sifts through the factors that can affect her MCNP runs' accuracy, but leaves the inappropriate thermal treatment completely out. In this work, the inaccuracy of the criticality calculations has been investigated carrying out sets of similar burn-up calculations differing from each other only in the applied thermal cross section sets. The widest discrepancies were detected between the results obtained applying the free gas model and those obtained applying the molecular models. This, in conjunction with the fact that the free gas model does not even keep in count the molecular structure of H{sub 2}O suggest to discard it and to focus the investigation on the vapour and liquid models. Dr. J. Marti, from the Universitat Politecnica de Catalunya, Barcelona, Spain registered the generalized frequency distributions obtained from the molecular dynamics simulations of 216 molecules of H{sub 2}O in 10 simulated supercritical states and published in an article (1999) the frequencies of the three characteristic distribution peaks for each simulated state, in numerical format. A confrontation with the corresponding peaks from Bernnat's available frequency distributions for liquid water and vapour revealed the peaks of the latter to be closest to the supercritical water ones in nearly all cases. Hence the inference that thermal cross section sets for vapour are for the time

  8. A study on the instrumentation technology for the irradiation of the DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Bae, K. K.; Moon, J. S.; Park, H. S.; Song, K. C.; Jung, I. H.; Kang, K. H.; Yang, M. S.; Ryu, J. S.; Cho, Y. G. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    The advanced countries have a great experience for the irradiation test of material and nuclear fuel from 1960's. Especially, instruments and instrument attachment technology and irradiation equipment have been developed for this purpose. The techniques described in this report are temperature measurement of fuel rod cladding surface and fuel centerline, cladding elongation, fuel stack elongation, fuel rod pressure by fission gas release. As a result of analysis we define the instrumentation techniques for the DUPIC fuel irradiation test at HANARO. The development of high temperature thermocouple, the method of attachment to fuel cladding and inside the fuel, the method of fuel rod elongation and fuel rod pressure measurement (by LVDT) are essential techniques. In addition to the instrumentation technique, re-instrumentation technique have been developed for high burn-up fuel in commercial reactor. The re-instrumentation is a technique of re-fabrication and instrumentation of irradiated fuel rod. It technology can be utilized as a reference for us to develop instrumented test rig with a remote manner. 19 refs., 38 figs. (Author)

  9. Utilization of the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kyung Bae; Chung, Yong Sam; Han, Hyun Soo and others

    1998-12-01

    Radioisotope facilities were operated to support research, development and production of radioisotopes. Irradiation facilities were operated to offer irradiation service to the clients. Even though intensive periodical checks and some essential maintenance had been carried out the results of RI facility maintenance seemed to be not much satisfactory due mainly to the shortage of skilled manpower and maintenance expenses. High intensity {gamma} irradiator was normally operated 426 times for overall 3,440 hrs, and the low intensity {gamma} irradiator was operated 157 times for overall 11,092 hrs, respectively. (author)

  10. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-22

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

  11. Manufacturing and characterisation of electrode membrane assemblies for low temperature fuel cells; Herstellung und Charakterisierung von Membran-Elektroden-Einheiten fuer Niedertemperatur Brennstoffzellen

    Energy Technology Data Exchange (ETDEWEB)

    Kaz, Till

    2008-08-22

    The high cost for a Polymer electrolyte Fuel Cell (PEFC) System is still a barrier for commercial breakthrough, which cannot be compensated by the advantages of being pollution free, or nearly noiseless. The most effective way of saving costs is to reduce expensive materials, because the material costs only for the Membrane Electrode Assemblies (MEAs) is more than 70% of the total costs of a PEFC Stack. Within the MEA a main part of the costs is due to the catalyst. It is one of the main goals to decrease the catalyst loading by simultaneously increasing the performance or keeping it at least constant. Because in most electrodes only 20-50% of the catalyst in the electrodes is used, enlarging the electrochemical active area is one of the key problems of the PEFC. For being electrochemical active, the catalyst must be reachable for the gases, he must have a good ionic conductivity to the membrane and he must be attached to the Gas Diffusion Layer (GDL) by electron conductivity. In literature often an inferior ionic contact of the catalyst to the membrane is responsible for the low catalyst utilization. In the first part of the work, model electrodes with different kinds of catalysts and different amounts of electrolyte in the electrodes were investigated to explore the interrelationship between platinum and electrolyte content. Three different catalysts, unsupported Pt- black, 60 wt.% Pt carbon-supported and 20 wt.% Pt carbon-supported with an addition of Nafion powder of 0%, 20%, 40%, 60 wt.%, and 80 wt.% were used. The electrodes were prepared by spraying the electrode material with the DLR dry spray technique directly onto the membrane and then rolling them while hot. Because material solutions were not used, the structure of the electrodes are determinable and predictable. Numerous different in- and ex-situ characterization methods like impedance spectroscopy, U-i characteristic, cyclic voltammetry, proton conductivity measurements, half-cell measurements and

  12. Turbine combustor with fuel nozzles having inner and outer fuel circuits

    Science.gov (United States)

    Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

    2013-12-24

    A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

  13. System and method for controlling a combustor assembly

    Science.gov (United States)

    York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Stevenson, Christian Xavier

    2013-03-05

    A system and method for controlling a combustor assembly are disclosed. The system includes a combustor assembly. The combustor assembly includes a combustor and a fuel nozzle assembly. The combustor includes a casing. The fuel nozzle assembly is positioned at least partially within the casing and includes a fuel nozzle. The fuel nozzle assembly further defines a head end. The system further includes a viewing device configured for capturing an image of at least a portion of the head end, and a processor communicatively coupled to the viewing device, the processor configured to compare the image to a standard image for the head end.

  14. Compliant fuel cell system

    Science.gov (United States)

    Bourgeois, Richard Scott [Albany, NY; Gudlavalleti, Sauri [Albany, NY

    2009-12-15

    A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

  15. Data summary report for the destructive examination of Rods G7, G9, J8, I9, and H6 from Turkey Point Fuel Assembly B17

    Energy Technology Data Exchange (ETDEWEB)

    Davis, R B; Pasupathi, V

    1981-04-01

    Destructive examination results of five spent fuel rods from a Turkey Point Unit 3 pressurized water reactor are reported. Examinations included fission gas analysis, cladding hydrogen content analysis, fuel burnup analysis, metallographic examination, autoradiography and shielded electron microprobe analysis. All rods were found to be of sound integrity with an average burnup of 27 GWd/MTU and a 0.3% fission gas release.

  16. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  17. Neutronic fuel element fabrication

    Science.gov (United States)

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  18. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  19. Molecular-Level Modeling of the Structure and Proton Transport within the Membrane Electrode Assembly of Hydrogen Proton Exchange Membrane Fuel Cells

    Science.gov (United States)

    Selvan, Myvizhi Esai; Keffer, David J.

    The creation of proton exchange membrane fuel cells (PEMFCs) in the early 1960's attracted great interest with the prospect of serving as a highly efficient and eco-friendly power source. This nascent technology found a broad range of applications spanning from spacecrafts to automobiles and electronic devices. The PEMFC in its simplest form consists of an anode, where the hydrogen fuel is catalytically electro-oxidized (dissociated into protons and electrons), a cathode, where oxygen is catalytically electro-reduced (combined with protons to form water) and a polymer electrolyte membrane, which serves as the structural framework of the cell and transports protons from anode to cathode, while the electrons are forced through the external circuit generating electricity. Today, fuel cell remains one of the most promising means of generating energy from alternative fuels, with tremendous potential to reduce oil dependence and carbon emissions. However, current PEMFCs have a relatively narrow operational range and a high cost of production, thus requiring significant experimental and theoretical research to develop a thorough understanding of this technology (at both the molecular and macroscopic scale), which will ultimately render the fuel cell as an economically viable option.

  20. The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Repetto, G. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Dominguez, C. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Durville, B. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Carnemolla, S. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Campello, D. [Institut National des Sciences Appliques, Lyon (France); Tardiff, N. [Institut National des Sciences Appliques, Lyon (France); Gradeck, M. [Univ. de Lorraine, Nancy, France. LEMTA

    2015-09-04

    The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the ballooned areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.

  1. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    Science.gov (United States)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  2. Arrangement and statistics of storage containers of spent fuel for assemblies of the SFP of NPP-L V, Unit 1; Arreglo y estadistica de contenedores de almacenamiento de combustible gastado para los ensambles de la ACG de la Unidad 1 de la Central Nucleoelectrica Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Vargas A, A. F.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    This work presents the determination of assemblies of the spent fuel pool (SFP) of the nuclear power plant of Laguna Verde (NPP-L V) which are candidates to be assigned to storage containers of independent spent fuel, with the objective of liberating decay heat and to have more space in the SFP, for the store of retired assemblies of the reactors in future reloads of NPP-L V, besides that the removed assemblies of the SFP should be stored in specific containers to guarantee the physical safety of them, as well as the radiological protection to the population and the environment. The design of the containers considered in this work is to store a maximum of 69 assemblies; it has a thermal capacity of 26 kilowatts and allows storing assemblies with a minimum of 5 years of have been extracted of the reactor core. Is considered that in 2016 start the storage of the spent assemblies on the containers, the candidates assemblies to store cover from the first reload in 1991, until the assemblies deposited in the SFP in the 14 reload in 2010; therefore in 2016, such assemblies will have fulfilled with the criteria of 5 years of have been removed of the Reactor, also the 69 assemblies assigned to each container will have a resulting decay heat that does not exceed the thermal capacity of the container, but that in great percentage approximates to the same one, and this way to take full advantage of their storage capacity and thermal capacity for each container. This work also contains the arrangement to accommodate the assemblies in the containers; such arrangement is constituted by areas according to the decay heat of each assembly. (Author)

  3. Pulverized fuel-oxygen burner

    Science.gov (United States)

    Taylor, Curtis; Patterson, Brad; Perdue, Jayson

    2017-09-05

    A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through the solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.

  4. 24 CFR 3285.601 - Field assembly.

    Science.gov (United States)

    2010-04-01

    ... 24 Housing and Urban Development 5 2010-04-01 2010-04-01 false Field assembly. 3285.601 Section... § 3285.601 Field assembly. Home manufacturers must provide specific installation instructions for the proper field assembly of manufacturer-supplied and shipped loose ducts, plumbing, and fuel supply system...

  5. Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013

    Energy Technology Data Exchange (ETDEWEB)

    SISKIND B.; N/A

    2013-06-03

    This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

  6. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  7. The transport of fuel assemblies. New containers for transport the used nuclear material in Juzbado factory; Como se transportan los elementos combustibles? Nuevos contenedores para transporte de material fisionable utilizados en la fabrica de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Juzbado Manufacturing Facility is designed to be versatile and flexible. It is manufactured different kind of fuel assemblies PWR, BWR and VVER, beginning by the uranium oxide coming from the conversion facilities. The transport of these products (radioactive material fissile) requires the availability of different kind of packages; our models variety is similar to the big manufacturers. It is required a depth knowledge of the licensing process, approvals, manufacturing and handling instruction to be confident. Moreover, the recently changes on the Transport Regulations and the demands for the approval by the Competent Authorities have required the renovation of most of the package designs for the transport of radioactive material fissile worldwide. ENUSA assumed time ago this renovation and it is nowadays in the pick moment of this process. If we also consider the complexity on the management of multimodal international transportations, the Logistic task for the transport of nuclear material associated to the Juzbado factory results in a real changeling area. (Author)

  8. Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US Army Research Laboratory Liquid-Fueled Thermophotovoltaic Power Source Demonstrator

    Science.gov (United States)

    2016-09-01

    The photovoltaic cell(s) then converts the thermal radiation to electrical energy, which can be delivered to a load or conditioning circuitry...isolation, and leverage thermal performance. This analysis was completed using Solid Works Flow Simulation, which can calculate either the steady...ARL-TR-7829 ● SEP 2016 US Army Research Laboratory Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US

  9. Analysis of platinum and fluorine in spent fuel cell membrane-electrode-assemblies; Die Analyse von Platin und Fluor in gebrauchten Brennstoffzellen-Membranelektroden

    Energy Technology Data Exchange (ETDEWEB)

    Voelker, M. [FEM Forschungsinstitut, Schwaebisch Gmuend (Germany)

    2007-01-15

    Methods for platinum analysis in spent fuel cell stacks have been compared. Extraction with copper gave highest yields, though standard deviations are not satisfying compared to analysis of automotive catalysts yet. Dissipation of platinum from membrane electrodes to other system components was found to be less than 1%. Fluorine potentially interfering with thermometallurgic platinum recovery was found in both the surveyed system types PEMFC and PAFC. (orig.)

  10. Methods of conditioning direct methanol fuel cells

    Science.gov (United States)

    Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

    2005-11-08

    Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer electrolyte membrane of the membrane electrode assembly to a cathode surface of the membrane electrode assembly, and an electrical current of polarity opposite to that in a functioning direct methanol fuel cell is drawn through the membrane electrode assembly, wherein methanol is oxidized at the cathode surface of the membrane electrode assembly while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

  11. Highly Stable Sr-Free Cobaltite-Based Perovskite Cathodes Directly Assembled on a Barrier-Layer-Free Y2 O3 -ZrO2 Electrolyte of Solid Oxide Fuel Cells.

    Science.gov (United States)

    Ai, Na; Li, Na; Rickard, William D A; Cheng, Yi; Chen, Kongfa; Jiang, San Ping

    2017-03-09

    Direct assembly is a newly developed technique in which a cobaltite-based perovskite (CBP) cathode can be directly applied to a barrier-layer-free Y2 O3 -ZrO2 (YSZ) electrolyte with no high-temperature pre-sintering steps. Solid oxide fuel cells (SOFCs) based on directly assembled CBPs such as La0.6 Sr0.4 Co0.2 Fe0.8 O3-δ show high performance initially but degrade rapidly under SOFC operation conditions at 750 °C owing to Sr segregation and accumulation at the electrode/electrolyte interface. Herein, the performance and interface of Sr-free CBPs such as LaCoO3-δ (LC) and Sm0.95 CoO3-δ (SmC) and their composite cathodes directly assembled on YSZ electrolyte was studied systematically. The LC electrode underwent performance degradation, most likely owing to cation demixing and accumulation of La on the YSZ electrolyte under polarization at 500 mA cm-2 and 750 °C. However, the performance and stability of LC electrodes could be substantially enhanced by the formation of LC-gadolinium-doped ceria (GDC) composite cathodes. Replacement of La by Sm increased the cell stability, and doping of 5 % Pd to form Sm0.95 Co0.95 Pd0.05 O3-δ (SmCPd) significantly improved the electrode activity. An anode-supported YSZ-electrolyte cell with a directly assembled SmCPd-GDC composite electrode exhibited a peak power density of 1.4 W cm-2 at 750 °C, and an excellent stability at 750 °C for over 240 h. The higher stability of SmC as compared to that of LC is most likely a result of the lower reactivity of SmC with YSZ. This study demonstrates the new opportunities in the design and development of intermediate-temperature SOFCs based on the directly assembled high-performance and durable Sr-free CBP cathodes. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  13. Spent fuel data for waste storage programs

    Energy Technology Data Exchange (ETDEWEB)

    Greene, E M

    1980-09-01

    Data on LWR spent fuel were compiled for dissemination to participants in DOE-sponsored waste storage programs. Included are mechanical descriptions of the existing major types of LWR fuel assemblies, spent LWR fuel fission product inventories and decay heat data, and inventories of LWR spent fuel currently in storage, with projections of future quantities.

  14. Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R. de

    1966-07-01

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs.

  15. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

    2009-02-23

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

  16. Cerenkov Characteristics of BWR Assemblies using a Prototype DCVD with a Back-Illuminated CCD. Prepared for the Canadian Safeguards Support Program and the Swedish Support Program

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.D.; Gerwing, A.F. [Channel Systems Inc., Pinawa MA (Canada); Maxwell, R. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); Larsson, M.; Axell, K.; Hildingsson, L. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Lindberg, B. [LENS-TECH AB, Skellefteaa (Sweden); Sundkvist, E. [Teleca Design and Development, Stockholm (Sweden)

    2003-11-01

    The Canadian and Swedish Safeguards Support Programs have developed a prototype Digital Cerenkov Viewing Device (DCVD) to verify spent fuel. Field measurements in Swedish nuclear power reactor fuel bays on BWR fuel and non-fuel assemblies resulted in new Cerenkov information that offers the possibility of computer-assisted verification of spent-fuel assemblies. A number of fuel assemblies with missing fuel rods were examined. The missing fuel rods are easily detected when not hidden under the lifting handle of the fuel assembly. Initial studies of off-angle viewing of these assemblies show promise for the detection of the missing fuel rods under the lifting handle. The quantitative nature of the charge-coupled device was examined. A number of procedures were developed to quantify parameters such as image intensity and alignment (collimation) of fuel and no fuel assemblies. The quantitative studies on fuel assembly intensity as a function of cooling time showed excellent agreement with the theoretical calculations.

  17. Proton exchange membrane fuel cells

    CERN Document Server

    Qi, Zhigang

    2013-01-01

    Preface Proton Exchange Membrane Fuel CellsFuel CellsTypes of Fuel CellsAdvantages of Fuel CellsProton Exchange Membrane Fuel CellsMembraneCatalystCatalyst LayerGas Diffusion MediumMicroporous LayerMembrane Electrode AssemblyPlateSingle CellStackSystemCell Voltage Monitoring Module (CVM)Fuel Supply Module (FSM)Air Supply Module (ASM)Exhaust Management Module (EMM)Heat Management Module (HMM)Water Management Module (WMM)Internal Power Supply Module (IPM)Power Conditioning Module (PCM)Communications Module (COM)Controls Module (CM)SummaryThermodynamics and KineticsTheoretical EfficiencyVoltagePo

  18. Sequence assembly

    DEFF Research Database (Denmark)

    Scheibye-Alsing, Karsten; Hoffmann, S.; Frankel, Annett Maria

    2009-01-01

    Despite the rapidly increasing number of sequenced and re-sequenced genomes, many issues regarding the computational assembly of large-scale sequencing data have remain unresolved. Computational assembly is crucial in large genome projects as well for the evolving high-throughput technologies...... and plays an important role in processing the information generated by these methods. Here, we provide a comprehensive overview of the current publicly available sequence assembly programs. We describe the basic principles of computational assembly along with the main concerns, such as repetitive sequences...... in genomic DNA, highly expressed genes and alternative transcripts in EST sequences. We summarize existing comparisons of different assemblers and provide a detailed descriptions and directions for download of assembly programs at: http://genome.ku.dk/resources/assembly/methods.html....

  19. MOX fuel arrangement for nuclear core

    Energy Technology Data Exchange (ETDEWEB)

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  20. Mox fuel arrangement for nuclear core

    Energy Technology Data Exchange (ETDEWEB)

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  1. MOX fuel arrangement for nuclear core

    Energy Technology Data Exchange (ETDEWEB)

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  2. MOX fuel arrangement for nuclear core

    Energy Technology Data Exchange (ETDEWEB)

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  3. Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Dionne, B. (Nuclear Engineering Division)

    2011-05-23

    To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D

  4. Active Interrogation for Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dougan, Arden [National Nuclear Security Administration (NNSA), Washington, DC (United States)

    2015-11-05

    The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

  5. PEM/SPE fuel cell

    Science.gov (United States)

    Grot, Stephen Andreas

    1998-01-01

    A PEM/SPE fuel cell including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates.

  6. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  7. Bacterial community shift and incurred performance in response to in situ microbial self-assembly graphene and polarity reversion in microbial fuel cell.

    Science.gov (United States)

    Chen, Junfeng; Zhang, Lihua; Hu, Yongyou; Huang, Wantang; Niu, Zhuyu; Sun, Jian

    2017-10-01

    In this work, bacterial community shift and incurred performance of graphene modified bioelectrode (GM-BE) in microbial fuel cell (MFC) were illustrated by high throughput sequencing technology and electrochemical analysis. The results showed that Firmicutes occupied 48.75% in graphene modified bioanode (GM-BA), while Proteobacteria occupied 62.99% in graphene modified biocathode (GM-BC), both were dominant bacteria in phylum level respectively. Typical exoelectrogens, including Geobacter, Clostridium, Pseudomonas, Geothrix and Hydrogenophaga, were counted 26.66% and 17.53% in GM-BA and GM-BC. GM-BE was tended to decrease the bacterial diversity and enrich the dominant species. Because of the enrichment of exoelectrogens and excellent electrical conductivity of graphene, the maximum power density of MFC with GM-BA and GM-BC increased 33.1% and 21.6% respectively, and the transfer resistance decreased 83.8% and 73.6% compared with blank bioelectrode. This study aimed to enrich the microbial study in MFC and broaden the development and application for bioelectrode. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. Air breathing direct methanol fuel cell

    Science.gov (United States)

    Ren, Xiaoming; Gottesfeld, Shimshon

    2002-01-01

    An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source. Water loss from the cell is minimized by making the conductive cathode assembly hydrophobic and the conductive anode assembly hydrophilic.

  9. HEDL contribution to SRL fuel recycle program. Quarterly report, January--March 1977. [Sensitivity analysis of LMFBR fuel fabrication cost

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, J.F.

    1977-08-01

    Research on LWR fuel cycle is being done in the following categories: economic studies (sensitivity analysis of LMFBR fuel fabrication costs), spent fuel receipt and storage (failure of PWR and BWR fuel assemblies), fuel materials preparation or finishing processes, reduction of TRU waste generation, and environmental impacts. 12 tables. (DLC)

  10. Galaxy And Mass Assembly (GAMA): Gas Fueling of Spiral Galaxies in the Local Universe. I. The Effect of the Group Environment on Star Formation in Spiral Galaxies

    Science.gov (United States)

    Grootes, M. W.; Tuffs, R. J.; Popescu, C. C.; Norberg, P.; Robotham, A. S. G.; Liske, J.; Andrae, E.; Baldry, I. K.; Gunawardhana, M.; Kelvin, L. S.; Madore, B. F.; Seibert, M.; Taylor, E. N.; Alpaslan, M.; Brown, M. J. I.; Cluver, M. E.; Driver, S. P.; Bland-Hawthorn, J.; Holwerda, B. W.; Hopkins, A. M.; Lopez-Sanchez, A. R.; Loveday, J.; Rushton, M.

    2017-03-01

    We quantify the effect of the galaxy group environment (for group masses of 1012.5-1014.0 M ⊙) on the current star formation rate (SFR) of a pure, morphologically selected sample of disk-dominated (I.e., late-type spiral) galaxies with redshift ≤0.13. The sample embraces a full representation of quiescent and star-forming disks with stellar mass M * ≥ 109.5 M ⊙. We focus on the effects on SFR of interactions between grouped galaxies and the putative intrahalo medium (IHM) of their host group dark matter halos, isolating these effects from those induced through galaxy-galaxy interactions, and utilizing a radiation transfer analysis to remove the inclination dependence of derived SFRs. The dependence of SFR on M * is controlled for by measuring offsets Δlog(ψ *) of grouped galaxies about a single power-law relation in specific SFR, {\\psi }* \\propto {M}* -0.45+/- 0.01, exhibited by non-grouped “field” galaxies in the sample. While a small minority of the group satellites are strongly quenched, the group centrals and a large majority of satellites exhibit levels of ψ * statistically indistinguishable from their field counterparts, for all M *, albeit with a higher scatter of 0.44 dex about the field reference relation (versus 0.27 dex for the field). Modeling the distributions in Δlog(ψ *), we find that (I) after infall into groups, disk-dominated galaxies continue to be characterized by a similar rapid cycling of gas into and out of their interstellar medium shown prior to infall, with inflows and outflows of ˜1.5-5 x SFR and ˜1-4 x SFR, respectively; and (II) the independence of the continuity of these gas flow cycles on M * appears inconsistent with the required fueling being sourced from gas in the circumgalactic medium on scales of ˜100 kpc. Instead, our data favor ongoing fueling of satellites from the IHM of the host group halo on ˜Mpc scales, I.e., from gas not initially associated with the galaxies upon infall. Consequently, the color

  11. Flashback resistant pre-mixer assembly

    Science.gov (United States)

    Laster, Walter R [Oviedo, FL; Gambacorta, Domenico [Oviedo, FL

    2012-02-14

    A pre-mixer assembly associated with a fuel supply system for mixing of air and fuel upstream from a main combustion zone in a gas turbine engine. The pre-mixer assembly includes a swirler assembly disposed about a fuel injector of the fuel supply system and a pre-mixer transition member. The swirler assembly includes a forward end defining an air inlet and an opposed aft end. The pre-mixer transition member has a forward end affixed to the aft end of the swirler assembly and an opposed aft end defining an outlet of the pre-mixer assembly. The aft end of the pre-mixer transition member is spaced from a base plate such that a gap is formed between the aft end of the pre-mixer transition member and the base plate for permitting a flow of purge air therethrough to increase a velocity of the air/fuel mixture exiting the pre-mixer assembly.

  12. Mixer Assembly for a Gas Turbine Engine

    Science.gov (United States)

    Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Fotache, Catalin G. (Inventor); Smith, Lance L. (Inventor); Hautman, Donald J. (Inventor)

    2015-01-01

    A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

  13. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  14. FAILED FUEL DISPOSITION STUDY

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.

    2004-12-20

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  15. Expertise on the Goesgen-Daeniken nuclear power plant on the granting of a licence for the construction and operation of a water storage pool for fuel assemblies at the site of the power plant; Gutachten zum Gesuch der Kernkraftwerk Goesgen-Daeniken AG um Erteilung der Bewilligung fuer den Bau und Betrieb eines Brennelement-Nasslagers auf dem Areal des Kernkraftwerks

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-04-15

    On June 26, 2002, the Goesgen-Daeniken AG nuclear power plant (KKG) delivered a request to the Swiss Federal Council for the granting of a licence for the construction and operation of a water storage pool for the on-site storage of the power plant's fuel assemblies. The present report contains the results of the examination of the request by the Federal Agency for the Safety of Nuclear Installations (HSK), to check that the projected storage pool satisfies the legal requirements from the point of view of nuclear safety and protection against radioactivity. A water storage pool already exists in the reactor building of KKG. It was conceived for a fuel cycle based on the reprocessing of the spent fuel assemblies. Its capacity is not sufficient when the spent fuel assemblies are no longer reprocessed but have to be transferred and stored in the Central Intermediate Storage Facility (ZWILAG) in Wuerenlingen because their heat production is too high. The capacity of the actual water pool allows a maximum cooling time of 5-6 years, while 7-10 years are required before transfer to ZWILAG. The projected new water storage pool has to be aircraft crash and earthquake proof, in the same way that the reactor building itself has to be. It can store a maximum of 1008 fuel assemblies. The water in the pool as well as the pool walls shield the radiation from of the fuel assemblies almost completely. Each fuel assembly is put into a square steel channel. The channel walls are lined with 6.11 mg/cm{sup 2} of the neutron absorbing nuclide B-10, which guaranties the subcriticality of the water pool even if the storage pool would be entirely filled with non-irradiated fuel assemblies with the maximal allowed enrichment or the maximal allowed content of Plutonium in case of MOX fuel assemblies, which is a very conservative assumption. The heat released by decay in the spent fuel assemblies is transferred to the pool water. Storage pool cooling is carried out by natural circulation

  16. Ambient pressure fuel cell system

    Science.gov (United States)

    Wilson, Mahlon S.

    2000-01-01

    An ambient pressure fuel cell system is provided with a fuel cell stack formed from a plurality of fuel cells having membrane/electrode assemblies (MEAs) that are hydrated with liquid water and bipolar plates with anode and cathode sides for distributing hydrogen fuel gas and water to a first side of each one of the MEAs and air with reactant oxygen gas to a second side of each one of the MEAs. A pump supplies liquid water to the fuel cells. A recirculating system may be used to return unused hydrogen fuel gas to the stack. A near-ambient pressure blower blows air through the fuel cell stack in excess of reaction stoichiometric amounts to react with the hydrogen fuel gas.

  17. Metallic fuel design development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

  18. Solid oxide fuel cell generator with removable modular fuel cell stack configurations

    Science.gov (United States)

    Gillett, J.E.; Dederer, J.T.; Zafred, P.R.; Collie, J.C.

    1998-04-21

    A high temperature solid oxide fuel cell generator produces electrical power from oxidation of hydrocarbon fuel gases such as natural gas, or conditioned fuel gases, such as carbon monoxide or hydrogen, with oxidant gases, such as air or oxygen. This electrochemical reaction occurs in a plurality of electrically connected solid oxide fuel cells bundled and arrayed in a unitary modular fuel cell stack disposed in a compartment in the generator container. The use of a unitary modular fuel cell stack in a generator is similar in concept to that of a removable battery. The fuel cell stack is provided in a pre-assembled self-supporting configuration where the fuel cells are mounted to a common structural base having surrounding side walls defining a chamber. Associated generator equipment may also be mounted to the fuel cell stack configuration to be integral therewith, such as a fuel and oxidant supply and distribution systems, fuel reformation systems, fuel cell support systems, combustion, exhaust and spent fuel recirculation systems, and the like. The pre-assembled self-supporting fuel cell stack arrangement allows for easier assembly, installation, maintenance, better structural support and longer life of the fuel cells contained in the fuel cell stack. 8 figs.

  19. Fossil fuels -- future fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  20. Cogasification of solid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Green, A.; Jurczyk, K.; Mullin, J. [Univ. of Florida, Gainesville, FL (United States); Zanardi, M. [Sao Paulo State Univ., Guaratingueta, Sao Paulo (Brazil). Energy Dept.; Peres, S. [Univ. of Pernambuco, Recife (Brazil)

    1996-12-31

    High efficiency gas turbine based systems, utility deregulation and more stringent environmental regulations strongly favor the use of natural gas over coal and other solid fuels in new electricity generators. Solid fuels could continue to compete, however, if a low cost gasifier fed by low cost feedstocks can be coupled with a gas turbine system. The authors examine on-site gasification of coal with other domestic fuels in an indirectly heated gasifier as a strategy to lower the costs of solid fuel systems. The systematics of gaseous pyrolysis yields assembled with the help of thermal measurement data and molecular models suggests blending carbonaceous fuels such as coal, coke or char with oxygenated fuels such as biomass, RDF, MSW, or dried sewage sludge. Such solid fuel blending can, with the help of inexpensive catalysts, achieve an optimum balance of volatiles, heating values and residual char thus reducing the technical demands upon the gasifier. Such simplifications should lower capital and operating costs of the gasifier to the mutual benefit of both solid fuel communities.

  1. Fuel services; Servicios de combustible

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

    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)

  2. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); LeDuc, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-25

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  3. Critical experiments with mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D.R. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1997-06-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er{sub 2}O{sub 3} at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs.

  4. General Assembly

    CERN Multimedia

    Staff Association

    2016-01-01

    5th April, 2016 – Ordinary General Assembly of the Staff Association! In the first semester of each year, the Staff Association (SA) invites its members to attend and participate in the Ordinary General Assembly (OGA). This year the OGA will be held on Tuesday, April 5th 2016 from 11:00 to 12:00 in BE Auditorium, Meyrin (6-2-024). During the Ordinary General Assembly, the activity and financial reports of the SA are presented and submitted for approval to the members. This is the occasion to get a global view on the activities of the SA, its financial management, and an opportunity to express one’s opinion, including taking part in the votes. Other points are listed on the agenda, as proposed by the Staff Council. Who can vote? Only “ordinary” members (MPE) of the SA can vote. Associated members (MPA) of the SA and/or affiliated pensioners have a right to vote on those topics that are of direct interest to them. Who can give his/her opinion? The Ordinary General Asse...

  5. Assembling consumption

    DEFF Research Database (Denmark)

    Assembling Consumption marks a definitive step in the institutionalisation of qualitative business research. By gathering leading scholars and educators who study markets, marketing and consumption through the lenses of philosophy, sociology and anthropology, this book clarifies and applies...... the investigative tools offered by assemblage theory, actor-network theory and non-representational theory. Clear theoretical explanation and methodological innovation, alongside empirical applications of these emerging frameworks will offer readers new and refreshing perspectives on consumer culture and market...... societies. This is an essential reading for both seasoned scholars and advanced students of markets, economies and social forms of consumption....

  6. General Assembly

    CERN Multimedia

    Staff Association

    2015-01-01

    Mardi 5 mai à 11 h 00 Salle 13-2-005 Conformément aux statuts de l’Association du personnel, une Assemblée générale ordinaire est organisée une fois par année (article IV.2.1). Projet d’ordre du jour : 1- Adoption de l’ordre du jour. 2- Approbation du procès-verbal de l’Assemblée générale ordinaire du 22 mai 2014. 3- Présentation et approbation du rapport d’activités 2014. 4- Présentation et approbation du rapport financier 2014. 5- Présentation et approbation du rapport des vérificateurs aux comptes pour 2014. 6- Programme 2015. 7- Présentation et approbation du projet de budget 2015 et taux de cotisation pour 2015. 8- Pas de modifications aux Statuts de l'Association du personnel proposée. 9- Élections des membres de la Commission é...

  7. General assembly

    CERN Multimedia

    Staff Association

    2015-01-01

    Mardi 5 mai à 11 h 00 Salle 13-2-005 Conformément aux statuts de l’Association du personnel, une Assemblée générale ordinaire est organisée une fois par année (article IV.2.1). Projet d’ordre du jour : Adoption de l’ordre du jour. Approbation du procès-verbal de l’Assemblée générale ordinaire du 22 mai 2014. Présentation et approbation du rapport d’activités 2014. Présentation et approbation du rapport financier 2014. Présentation et approbation du rapport des vérificateurs aux comptes pour 2014. Programme 2015. Présentation et approbation du projet de budget 2015 et taux de cotisation pour 2015. Pas de modifications aux Statuts de l'Association du personnel proposée. Élections des membres de la Commission électorale. &am...

  8. General Assembly

    CERN Multimedia

    Staff Association

    2016-01-01

    Mardi 5 avril à 11 h 00 BE Auditorium Meyrin (6-2-024) Conformément aux statuts de l’Association du personnel, une Assemblée générale ordinaire est organisée une fois par année (article IV.2.1). Projet d’ordre du jour : Adoption de l’ordre du jour. Approbation du procès-verbal de l’Assemblée générale ordinaire du 5 mai 2015. Présentation et approbation du rapport d’activités 2015. Présentation et approbation du rapport financier 2015. Présentation et approbation du rapport des vérificateurs aux comptes pour 2015. Programme de travail 2016. Présentation et approbation du projet de budget 2016 Approbation du taux de cotisation pour 2017. Modifications aux Statuts de l'Association du personnel proposée. Élections des membres de la Commissio...

  9. General Assembly

    CERN Multimedia

    Staff Association

    2017-01-01

    Conformément aux statuts de l’Association du personnel, une Assemblée générale ordinaire est organisée une fois par année (article IV.2.1). Projet d’ordre du jour : Adoption de l’ordre du jour. Approbation du procès-verbal de l’Assemblée générale ordinaire du 5 avril 2016. Présentation et approbation du rapport d’activités 2016. Présentation et approbation du rapport financier 2016. Présentation et approbation du rapport des vérificateurs aux comptes pour 2016. Programme de travail 2017. Présentation et approbation du projet de budget 2017 Approbation du taux de cotisation pour 2018. Modifications aux Statuts de l'Association du personnel proposées. Élections des membres de la Commission électorale. Élections des vérifica...

  10. Fuel cell membrane humidification

    Science.gov (United States)

    Wilson, Mahlon S.

    1999-01-01

    A polymer electrolyte membrane fuel cell assembly has an anode side and a cathode side separated by the membrane and generating electrical current by electrochemical reactions between a fuel gas and an oxidant. The anode side comprises a hydrophobic gas diffusion backing contacting one side of the membrane and having hydrophilic areas therein for providing liquid water directly to the one side of the membrane through the hydrophilic areas of the gas diffusion backing. In a preferred embodiment, the hydrophilic areas of the gas diffusion backing are formed by sewing a hydrophilic thread through the backing. Liquid water is distributed over the gas diffusion backing in distribution channels that are separate from the fuel distribution channels.

  11. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  12. Fuel consolidation demonstration: Consolidation concept development

    Energy Technology Data Exchange (ETDEWEB)

    1990-02-01

    EPRI, Northeast utilities Service Company (NUSCO), DOE, Baltimore Gas Electric Company, and Combustion Engineering, Inc. (C-E) are engaged in a program to develop a system for consolidating spent fuel, in which the consolidated fuel will be licensable by NRC for storage in the spent-fuel storage pool. Fuel consolidation offers a means of substantially increasing the capacity of spent-fuel storage pools. Consolidation equipment design, development, construction, and testing are being performed by C-E in Windsor, Connecticut. Seismic and structural evaluation of the capability of the Millstone Unit 2 spent-fuel pool and building to accommodate the increased fuel capacity is being conducted by NUSCO. NUSCO plans to obtain a license to store consolidated fuel in the Millstone-2 spent-fuel storage pool. NUSCO also plans to perform a hot demonstration of the integrated consolidation system with spent fuel at Millstone-2. This report describes the consolidation system design that forms the basis for the detailed design of the equipment comprising the system, including information on the fabrication and testing of the equipment. Appendix B describes an evaluation of the ability of the system under development to consolidate LWR spent-fuel assemblies other than the 14 {times} 14 fuel of C-E design stored at Millstone-2. A comparison was made of fuel-assembly designs on the basis of information available in open literature. It was concluded that with appropriate dimensional modifications the spent-fuel consolidation system equipment design is applicable to almost all PWR fuel-assembly configurations. 8 refs., 20 figs.

  13. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  14. Method of improving fuel cell performance by removing at least one metal oxide contaminant from a fuel cell electrode

    Science.gov (United States)

    Kim, Yu Seung [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Zelenay, Piotr [Los Alamos, NM

    2009-08-18

    A method of removing contaminants from a fuel cell catalyst electrode. The method includes providing a getter electrode and a fuel cell catalyst electrode having at least one contaminant to a bath and applying a voltage sufficient to drive the contaminant from the fuel cell catalyst electrode to the getter electrode. Methods of removing contaminants from a membrane electrode assembly of a fuel cell and of improving performance of a fuel cell are also provided.

  15. Chalcogen catalysts for polymer electrolyte fuel cell

    Science.gov (United States)

    Alonso-Vante, Nicolas [Buxerolles, FR; Zelenay, Piotr [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Wieckowski, Andrzej [Champaign, IL; Cao, Dianxue [Urbana, IL

    2009-09-15

    A methanol-tolerant cathode catalyst and a membrane electrode assembly for fuel cells that includes such a cathode catalyst. The cathode catalyst includes a support having at least one transition metal in elemental form and a chalcogen disposed on the support. Methods of making the cathode catalyst and membrane electrode assembly are also described.

  16. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  17. Development of a membrane electrode as assembly production process for proton exchange membrane fuel cell (PEMFC) by sieve printing; Desenvolvimento de processo de producao de conjuntos eletrodo-membrana-eletrodo para celulas a combustivel baseadas no uso de membrana polimerica conditora de protons (PEMFC) por impressa a tela

    Energy Technology Data Exchange (ETDEWEB)

    Bonifacio, Rafael Nogueira

    2010-07-01

    Energy is a resource that presents historical trend of growth in demand. Projections indicate that future energy needs will require a massive use of hydrogen as fuel. The use of systems based on the use of proton exchange membrane fuel cell (PEMFC) has features that allow its application for stationary applications, automotive and portable power generation. The use of hydrogen as fuel for PEMFC has the advantage low pollutants' emission, when compared to fossil fuels. For the reactions in a PEMFC is necessary to build membrane electrode assembly (MEA). And the production of MEAs and its materials are relevant to the final cost of k W of power generated by systems of fuel cell. This represent currently a technological and financial barriers to large-scale application of this technology. In this work a process of MEAs fabrication were developed that showed high reproducibility, rapidity and low cost by sieve printing. The process of sieve printing and the ink composition as a precursor to the catalyst layer were developed, which allow the preparation of electrodes for MEAs fabrication with the implementation of the exact catalyst loading, 0.6 milligrams of platinum per square centimeters (mgPt.cm{sup -2}) suitable for cathodes and 0.4 mgPt.cm{sup -2} for anode in only one application step per electrode. The ink was developed, produced, characterized and used with similar characteristics to ink of sieve printing build for other applications. The MEAs produced had a performance of up to 712 m A.cm{sup -2} by 600 mV to 25 cm{sup 2} MEA area. The MEA cost production for MEAs of 247.86 cm{sup 2}, that can generate 1 kilowatt of energy was estimated to US$ 7,744.14 including cost of equipment, materials and labor. (author)

  18. Consolidation and disposal of PWR fuel inserts

    Energy Technology Data Exchange (ETDEWEB)

    Wakeman, B.H. (Virginia Electric and Power Co., Glen Allen, VA (United States))

    1992-08-01

    Design and licensing of the Surry Power Station Independent Spent Fuel Storage Installation was initiated in 1982 by Virginia Power as part of a comprehensive strategy to increase spent fuel storage capacity at the Station. Designed to use large, metal dry storage casks, the Surry Installation will accommodate 84 such casks with a total storage capacity of 811 MTU of spent pressurized water reactor fuel assemblies. Virginia Power provided three storage casks for testing at the Idaho National Engineerinq Laboratory's Test Area North and the testing results have been published by the Electric Power Research Institute. Sixty-nine spent fuel assemblies were transported in truck casks from the Surry Power Station to Test Area North for testing in the three casks. Because of restrictions imposed by the cask testing equipment at Test Area North, the irradiated insert components stored in these fuel assemblies at Surry were removed prior to transport of the fuel assemblies. Retaining these insert components proved to be a problem because of a shortage of spent fuel assemblies in the spent fuel storage pool that did not already contain insert components. In 1987 Virginia Power contracted with Chem-Nuclear Systems, Inc. to process and dispose of 136 irradiated insert components consisting of 125 burnable poison rod assemblies, 10 thimble plugging devices and 1 part-length rod cluster control assembly. This work was completed in August and September 1987, culminating in the disposal at the Barnwell, SC low-level radioactive waste facility of two CNS 3-55 liners containing the consolidated insert components.

  19. Technical Development of Gamma Scanning for Irradiated Fuel Rod after Upgrade of System in Hot-cell

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, Hee Moon; Baik, Seung Je; Yoo, Byung Ok; Choo, Yong Sun

    2007-06-15

    Non-destructive test system was installed at hot-cell(M1) in IMEF(Irradiated Materials Examination Facility) more than 10 years ago for the diametric measurement and gamma scanning of fuel rod. But this system must be needed to be remodeled for the effective operations. In 2006, the system was upgraded for 3 months. The collimator bench can be movable with horizontal direction(x-direction) by motorized system for sectional gamma scanning and 3-dimensional tomography of fuel rod. So, gamma scanning for fuel rod can be detectable by x, y and rotation directions. It may be possible to obtain the radioactivities with radial and axial directions of pellet. This system is good for the series experiments with several positions. Operation of fuel bench and gamma detection program were linked each other by new program tools. It can control detection and bench moving automatically when gamma inspection of fuel rod is carried out with axial or radial positions. Some of electronic parts were added in PLC panel, and operating panel was re-designed for the remote control. To operate the fuel bench by computer, AD converter and some I/O cards were installed in computer. All of software were developed in Windows-XP system instead of DOS system. Control programs were made by visual-C language. After upgrade of system, DUPIC fuel which was irradiated in HANARO research reactor was detected by gamma scanning. The results were good and operation of gamma scanning showed reduced inspection time and easy control of data on series of detection with axial positions. With consideration of ECT(Eddy Current Test) installation, the computer program and hardware were set up as well. But ECT is not installed yet, so we have to check abnormal situation of program and hardware system. It is planned to install ECT in 2007.

  20. Upgraded HFIR Fuel Element Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Sease, John D [ORNL

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  1. Fuel cell membranes and crossover prevention

    Science.gov (United States)

    Masel, Richard I [Champaign, IL; York, Cynthia A [Newington, CT; Waszczuk, Piotr [White Bear Lake, MN; Wieckowski, Andrzej [Champaign, IL

    2009-08-04

    A membrane electrode assembly for use with a direct organic fuel cell containing a formic acid fuel includes a solid polymer electrolyte having first and second surfaces, an anode on the first surface and a cathode on the second surface and electrically linked to the anode. The solid polymer electrolyte has a thickness t:.gtoreq..times..times..times..times. ##EQU00001## where C.sub.f is the formic acid fuel concentration over the anode, D.sub.f is the effective diffusivity of the fuel in the solid polymer electrolyte, K.sub.f is the equilibrium constant for partition coefficient for the fuel into the solid polymer electrolyte membrane, I is Faraday's constant n.sub.f is the number of electrons released when 1 molecule of the fuel is oxidized, and j.sub.f.sup.c is an empirically determined crossover rate of fuel above which the fuel cell does not operate.

  2. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  3. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  4. Optical fuel pin scanner. [Patent application; for reading identifications

    Science.gov (United States)

    Kirchner, T.L.; Powers, H.G.

    1980-12-09

    This patent relates to an optical identification system developed for post-irradiation disassembly and analysis of fuel bundle assemblies. The apparatus is designed to be lowered onto a stationary fuel pin to read identification numbers or letters imprinted on the circumference of the top fuel pin and cap. (DLC)

  5. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  6. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  7. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  8. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  9. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  10. Sputter-deposited fuel cell membranes and electrodes

    Science.gov (United States)

    Narayanan, Sekharipuram R. (Inventor); Jeffries-Nakamura, Barbara (Inventor); Chun, William (Inventor); Ruiz, Ron P. (Inventor); Valdez, Thomas I. (Inventor)

    2001-01-01

    A method for preparing a membrane for use in a fuel cell membrane electrode assembly includes the steps of providing an electrolyte membrane, and sputter-depositing a catalyst onto the electrolyte membrane. The sputter-deposited catalyst may be applied to multiple sides of the electrolyte membrane. A method for forming an electrode for use in a fuel cell membrane electrode assembly includes the steps of obtaining a catalyst, obtaining a backing, and sputter-depositing the catalyst onto the backing. The membranes and electrodes are useful for assembling fuel cells that include an anode electrode, a cathode electrode, a fuel supply, and an electrolyte membrane, wherein the electrolyte membrane includes a sputter-deposited catalyst, and the sputter-deposited catalyst is effective for sustaining a voltage across a membrane electrode assembly in the fuel cell.

  11. Bos taurus genome assembly.

    Science.gov (United States)

    Liu, Yue; Qin, Xiang; Song, Xing-Zhi Henry; Jiang, Huaiyang; Shen, Yufeng; Durbin, K James; Lien, Sigbjørn; Kent, Matthew Peter; Sodeland, Marte; Ren, Yanru; Zhang, Lan; Sodergren, Erica; Havlak, Paul; Worley, Kim C; Weinstock, George M; Gibbs, Richard A

    2009-04-24

    We present here the assembly of the bovine genome. The assembly method combines the BAC plus WGS local assembly used for the rat and sea urchin with the whole genome shotgun (WGS) only assembly used for many other animal genomes including the rhesus macaque. The assembly process consisted of multiple phases: First, BACs were assembled with BAC generated sequence, then subsequently in combination with the individual overlapping WGS reads. Different assembly parameters were tested to separately optimize the performance for each BAC assembly of the BAC and WGS reads. In parallel, a second assembly was produced using only the WGS sequences and a global whole genome assembly method. The two assemblies were combined to create a more complete genome representation that retained the high quality BAC-based local assembly information, but with gaps between BACs filled in with the WGS-only assembly. Finally, the entire assembly was placed on chromosomes using the available map information.Over 90% of the assembly is now placed on chromosomes. The estimated genome size is 2.87 Gb which represents a high degree of completeness, with 95% of the available EST sequences found in assembled contigs. The quality of the assembly was evaluated by comparison to 73 finished BACs, where the draft assembly covers between 92.5 and 100% (average 98.5%) of the finished BACs. The assembly contigs and scaffolds align linearly to the finished BACs, suggesting that misassemblies are rare. Genotyping and genetic mapping of 17,482 SNPs revealed that more than 99.2% were correctly positioned within the Btau_4.0 assembly, confirming the accuracy of the assembly. The biological analysis of this bovine genome assembly is being published, and the sequence data is available to support future bovine research.

  12. Bos taurus genome assembly

    Directory of Open Access Journals (Sweden)

    Sodergren Erica

    2009-04-01

    Full Text Available Abstract Background We present here the assembly of the bovine genome. The assembly method combines the BAC plus WGS local assembly used for the rat and sea urchin with the whole genome shotgun (WGS only assembly used for many other animal genomes including the rhesus macaque. Results The assembly process consisted of multiple phases: First, BACs were assembled with BAC generated sequence, then subsequently in combination with the individual overlapping WGS reads. Different assembly parameters were tested to separately optimize the performance for each BAC assembly of the BAC and WGS reads. In parallel, a second assembly was produced using only the WGS sequences and a global whole genome assembly method. The two assemblies were combined to create a more complete genome representation that retained the high quality BAC-based local assembly information, but with gaps between BACs filled in with the WGS-only assembly. Finally, the entire assembly was placed on chromosomes using the available map information. Over 90% of the assembly is now placed on chromosomes. The estimated genome size is 2.87 Gb which represents a high degree of completeness, with 95% of the available EST sequences found in assembled contigs. The quality of the assembly was evaluated by comparison to 73 finished BACs, where the draft assembly covers between 92.5 and 100% (average 98.5% of the finished BACs. The assembly contigs and scaffolds align linearly to the finished BACs, suggesting that misassemblies are rare. Genotyping and genetic mapping of 17,482 SNPs revealed that more than 99.2% were correctly positioned within the Btau_4.0 assembly, confirming the accuracy of the assembly. Conclusion The biological analysis of this bovine genome assembly is being published, and the sequence data is available to support future bovine research.

  13. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

    ... spent fuel (i.e., intact assembly or consolidated fuel rods), the inerting atmosphere requirements. (b... maintained in a subcritical condition under credible conditions. (d) Radiation shielding and confinement...

  14. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising