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Sample records for fuel assembly designs

  1. 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)

  2. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. Most advanced HTP fuel assembly design for EPR

    International Nuclear Information System (INIS)

    Francillon, Eric; Kiehlmann, Horst-Dieter

    2006-01-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  5. Fuel assembly bow: analytical modeling and resulting design improvements

    International Nuclear Information System (INIS)

    Stabel, J.; Huebsch, H.P.

    1995-01-01

    The bowing of fuel assemblies may result in a contact between neighbouring fuel assemblies and in connection with a vibration to a resulting wear or even perforation at the corners of the spacer grids of neighbouring assemblies. Such events allowed reinsertion of a few fuel assemblies in Germany only after spacer repair. In order to identify the most sensitive parameters causing the observed bowing of fuel assemblies a new computer model was develop which takes into a account the highly nonlinear behaviour of the interaction between fuel rods and spacers. As a result of the studies performed with this model, design improvements such as a more rigid connection between guide thimbles and spacer grids, could be defined. First experiences with this improved design show significantly better fuel behaviour. (author). 5 figs., 1 tabs

  6. Mixed Reload Design Using MOX and UOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Ramon, Ramirez Sanchez J.; Perry, R.T.

    2002-01-01

    As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

  7. ABB. CASE's GUARDIANTM Debris Resistant Fuel Assembly Design

    International Nuclear Information System (INIS)

    Dixon, D. J.; Wohlsen, W. D.

    1992-01-01

    ABB CE's experience, that 72% of all recent fuel-rod failures are caused by debris fretting, is typical. In response to this problem, ABB Combustion Engineering began supplying in the late 1980s fuel assemblies with a variety of debris resistant features, including both long-end caps and small flow holes. Now ABB CAE has developed an advanced debris resistant design concept, GUARDIAN TM , which has the advantage of capturing and retaining more debris than other designs, while displacing less plenum or active fuel volume than the long end-cap design. GUARDIAN TM design features have now been implemented into four different assembly designs. ABB CASE's GUARDIAN TM fuel assembly is an advanced debris-resistant design which has both superior filtering performance and uniquely, excellent debris retention, Retention effectively removes the debris from circulation in the coolant so that it is not able to threaten the fuel again. GUARDIAN TM features have been incorporated into four ABB. CAE fuel assembly designs. These assemblies are all fully compatible with the NSLS, and full-batch operation with GUARDIAN TM began in 1992. The number of plants of both CAE and non-CAE design which accept GUARDIAN TM for debris protection is expected to grow significantly during the next few years

  8. 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.

  9. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. 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.

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  13. Overview of neutronic fuel assembly design and in-core fuel management

    International Nuclear Information System (INIS)

    Porsch, D.; Charlier, A.; Meier, G.; Mougniot, J.C.; Tsuda, K.

    2000-01-01

    The civil and military utilization of nuclear power results in stockpiles of spent fuel and separated plutonium. Recycling of the recovered plutonium in Light Water Reactors (LWR) is currently practiced in Belgium, France, Germany, and Switzerland, in Japan it is in preparation. Modern MOX fuel, with its optimized irradiation and reprocessing behavior, was introduced in 1981. Since then, about 1700 MOX fuel assemblies of different mechanical and neutronic design were irradiated in commercial LWRs and reached fuel assembly averaged exposures of up to 51.000 MWd/t HM. MOX fuel assemblies reloaded in PWR have an average fissile plutonium content of up to 4.8 w/o. For BWR, the average fissile plutonium content in actual reloads is 3.0 w/o. Targets for the MOX fuel assembly design are the compatibility to uranium fuel assemblies with respect to their mechanical fuel rod and fuel assembly design, they should have no impact on the flexibility of the reactor operation, and its reload should be economically feasible. In either cycle independent safety analyses or individually for each designed core it has to be demonstrated that recycling cores meet the same safety criteria as uranium cores. The safety criteria are determined for normal operation and for operational as well as design basis transients. Experience with realized MOX core loadings confirms the reliability of the applied modern design codes. Studies for reloads of advanced MOX assemblies in LWRs demonstrate the feasibility of a future development of the thermal plutonium recycling. New concepts for the utilization of plutonium are under consideration and reveal an attractive potential for further developments on the plutonium exploitation sector. (author)

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

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    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.)

  15. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakamura, Mitsuya; Yamashita, Jun-ichi; Mochida, Takaaki.

    1986-01-01

    Purpose: To improve the fuel economy by increasing the reactivity at the latter burning stage of fuel assemblies and thereby increasing the burn-up degree. Constitution: At the later stage of the burning where the infinite multiplication factor of a fuel assembly is lowered, fuel rods are partially discharged to increase the fuel-moderator volume ratio in the fuel assembly. Then, plutonium is positively burnt by bringing the ratio near to an optimum point where the infinite multiplication factor becomes maximum and the reactivity of the fuel assembly is increased by utilizing the spectral shift effect. The number of the fuel rods to be removed is selected so as to approach the fuel-moderator atom number ratio where the infinite multiplication factor is maximum. Further, the positions where the thermal neutron fluxes are low are most effective for removing the rods and those positions between which no fuel rods are present and which are adjacent with neither the channel box nor the water rods are preferred. The rods should be removed at the time when the burning is proceeded at lest for one cycle. The reactivity is thus increased and the burn-up degree of fuels upon taking-out can be improved. (Kamimura, M.)

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  18. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  2. New requirements for the WWER fuel and their consideration in designing the fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Ananyev, Y.

    2003-01-01

    In 2001-2002 the base designs of the new generation fuel assemblies for the WWER-440 and WWER-1000 reactors were developed. The ways of their further modernisation were defined. The present report deals with the urgent requirements and how they have been implemented in these designs. The assessment of the efficiency of new designs is carried out on the basis of the existing data of the world market on the cost of: Uranium concentrate; dividing operations; fabrication. It is additionally possible also to take into account the cost of transportation, storage and processing of the irradiated fuel including burial of wastes

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  4. Design, fabrication and in-core management of advanced fuel assemblies for LWRs

    International Nuclear Information System (INIS)

    Schmiedel, P.

    1987-01-01

    For the operation of a nuclear power plant the most important requirements concerning fuel assembly supply are: 1. reliability and flexibility; 2. operational safety; 3. optimum economy. In order to achieve this the fuel assembly suppliers have to harmonize the individual functions in their organization, i.e. R and D effeorts, design and manufacturing as well as quality control and assurance measurements, in-core fuel management and fuel assembly services. The status of the technical progress regarding morden KWU fuel assemblies for LWRs are summarized. (Liu)

  5. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  6. Modernization of the design and optimization of the manufacturing technology of RBMK fuel rods and fuel assemblies

    International Nuclear Information System (INIS)

    Panushkin, A.K.; Tsiboulia, V.A.; Bek, E.

    1998-01-01

    The paper describes design and experience in fabrication of fuel and fuel channels for RBMK reactors (RBMK-1000 and RBMK-1500) at the JSC ''Mashinostroitelny Zavod'', Electrostal, Russia. The most important measures developed and undertaken since Chernobyl accident to increase operational safety of RBMK reactors are presented. Emphasis is given to modifications in fuel design and technology including U-Er fuel, rods with a new plug and fuel assemblies with Zr spacer grids. (author)

  7. Improvement of the design model for SMART fuel assembly

    International Nuclear Information System (INIS)

    Zee, Sung Kyun; Yim, Jeong Sik

    2001-04-01

    A Study on the design improvement of the TEP, BEP and Hoddown spring of a fuel assembly for SMART was performed. Cut boundary Interpolation Method was applied to get more accurate results of stress and strain distribution from the results of the coarse model calculation. The improved results were compared with that of a coarse one. The finer model predicted slightly higher stress and strain distribution than the coarse model, which meant the results of the coarse model was not converged. Considering that the test results always showed much less stress than the FEM and the location of the peak stress of the refined model, the pressure stress on the loading point seemed to contribute significantly to the stresses. Judging from the fact that the peak stress appeared only at the local area, the results of the refined model were considered enough to be a conservative prediction of the stress levels. The slot of the guide thimble screw was ignored to get how much thickness of the flow plate can be reduced in case of optimization of the thickness and also cut off the screw dent hole was included for the actual geometry. For the BEP, the leg and web were also included in the model and the results with and without the leg alignment support were compared. Finally, the holddown spring which is important during the in-reactor behavior of the FA was modeled more realistic and improved to include the effects of the friction between the leaves and the loading surface. Using this improved model, it was possible that the spring characteristics were predicted more accurate to the test results. From the analysis of the spring characteristics, the local plastic area controled the characteristics of the spring dominantly which implied that it was necessary for the design of the leaf to be optimized for the improvement of the plastic behavior of the leaf spring

  8. Nuclear fuel assembly repair

    International Nuclear Information System (INIS)

    Bassler, E.A.; Stavsky, R.

    1986-01-01

    In response to utility needs to recover investment in nuclear fuel assemblies, Westinghouse Electric Corporation has developed tools and equipment to repair damaged fuel assemblies in an economical and safe manner, to enable utilities to reinsert these assemblies in the core. There are two possible repair techniques - bottom nozzle reconstitution and top nozzle reconstitution. Both techniques have been approved through formal design review; prototype tools have been built and successfully tested. The tools are modular in nature, easily transportable, and designed to fit the spent fuel pool at a reactor site. (author)

  9. Assembly Bow Characteristics of the HIPER16TM Fuel Design

    International Nuclear Information System (INIS)

    Jeon, Sang-Youn; Kwon, O-Cheol; Ha, Dong-Geun; Kim, Jae-Ik

    2015-01-01

    The out-of-pile tests were performed either in air or in a hydraulic loop and at room temperature or operating temperature conditions. The test results include the required physical and thermal-hydraulic data needed to verify the HIPER16 TM fuel design. The mechanical integrity and safety of HIPER16 TM fuel design has been verified based on the final verification tests and evaluations. The visual examinations and dimensional measurements were performed on the LTAs using poolside examination equipment. The in-reactor verification test results showed that the HIPER16 TM fuel design met the irradiation related design requirement. The poolside examinations after 3rd irradiation cycle of LTA will be performed in the end of 2015.

  10. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  11. Hydraulic Design of the CARA Fuel Assembly for Atucha-I

    International Nuclear Information System (INIS)

    Juanico, Luis; Brasnarof, Daniel

    2000-01-01

    In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

  12. A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Bessho, Yasunori; Ishii, Yoshihiko; Sadaoka, Noriyuki.

    1990-01-01

    Burnable poisons are disposed in the lower portions of a water rod, a channel box and a control rod guide pipe in a fuel assembly, and the amount for each of them is set to burn out in one operation cycle. Since the inner side of the water rod and the control rod guide pipe and gaps are filled with steams at the initial and the intermediate stages of the operation cycle, moderation of neutrons is delayed to harden the spectrum. On the other hand, since the burnable poisons are burnt out in the final stage of the operation cycle, γ-ray heating is not expected and since the insides of the water rod and the control rod guide pipe and the gaps are filled with water of great moderation effect, the neutron spectrum arae softened. In view of the above, void coefficient is increased to promote conversion from U-235 to Pu-239 by utilizing exothermic reaction of burnable poisons at the initial and the intermediate stages in the operation cycle and generation of voids are eliminated at the final stage where the burnable poisons are burnt out, thereby enabling effective burning of Pu-239. (N.H.)

  14. A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

    International Nuclear Information System (INIS)

    Aparicio, Gaspar; Di Marco, Agustin; Falcone, Jose M.; Giorgis, Miguel A.; Mathot, Sergio R.; Migliori, Julio; Orlando, Oscar S.; Restelli, Miguel A.; Ruggirello, Gabriel; Sapia, Gustavo C.; Zinzallari, Fausto; Bianchi, Daniel R.; Volpi, Ricardo M.

    2000-01-01

    The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

  15. RCC-C: Design and construction rules for fuel assemblies of PWR nuclear power plants

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-C code contains all the requirements for the design, fabrication and inspection of nuclear fuel assemblies and the different types of core components (rod cluster control assemblies, burnable poison rod assemblies, primary and secondary source assemblies and thimble plug assemblies). The design, fabrication and inspection rules defined in RCC-C leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build nuclear fuel assemblies and incorporate the resulting feedback. The code's scope covers: fuel system design, especially for assemblies, the fuel rod and associated core components, the characteristics to be checked for products and parts, fabrication methods and associated inspection methods. The RCC-C code is used by the operator of the PWR nuclear power plants in France as a reference when sourcing fuel from the world's top two suppliers in the PWR market, given that the French operator is the world's largest buyer of PWR fuel. Fuel for EPR projects is manufactured according to the provisions of the RCC-C code. The code is available in French and English. The 2005 edition has been translated into Chinese. Contents of the 2015 edition of the RCC-C code: Chapter 1 - General provisions: 1.1 Purpose of the RCC-C, 1.2 Definitions, 1.3 Applicable standards, 1.4 Equipment subject to the RCC-C, 1.5 Management system, 1.6 Processing of non-conformances; Chapter 2 - Description of the equipment subject to the RCC-C: 2.1 Fuel assembly, 2.2 Core components; Chapter 3 - Design: Safety functions, operating functions and environment of fuel assemblies and core components, design and safety principles; Chapter 4 - Manufacturing: 4.1 Materials and part characteristics, 4.2 Assembly requirements, 4.3 Manufacturing and inspection processes, 4.4 Inspection methods, 4.5 Certification of NDT inspectors, 4.6 Characteristics to be inspected for the

  16. Fuel assemblies

    International Nuclear Information System (INIS)

    Sadaoka, Noriyuki.

    1986-01-01

    Purpose: To maintain a satisfactory integrity by preventing the increase of corrosion at the outer surface of a fuel can near the point of contact between the fuel can and the spacer due to the use of fuel pellets incorporated with burnable poisons. Constitution: Since reactor coolants are at high temperature and high pressure, zirconium and water are brought into reaction to proceed oxidation at the outer surface of a fuel can to form uniform oxidation layers. However, abrasion corrosion is additionally formed at the contact portion between the spacer and the fuel can, by which the corrosion is increased by about 25 %. For preventing such nodular corrosion, fuel pellets not incorporated with burnable poisons are charged at a portion of the fuel rod where the spacer is supported and fuel pellets incorporated with burnable poisons are charged at the positions other than about to thereby suppress the amount of the corrosion at the portion where the corrosion of the fuel can is most liable to be increased to thereby improve the fuel integrity. That is, radiolysis of coolants due to gamma-rays produced from gadolinium is lowered to reduce the oxygen concentration near the outer surface thereby preventing the corrosion. (Kawakami, Y.)

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitosi.

    1993-01-01

    Fuel pellets containing burnable poison and fuel pellets not containing burnable poison are used together in burnable poison-incorporated fuel rods which is disposed at the outermost layer of a cluster. Since the burnable poison-incorporated fuel rods are disposed at the outermost layer of the cluster where a neutron flux level is high and, accordingly, the power is high originally, local power peaking can be suppressed and, simultaneously, fuels can be burnt effectively without increasing the fuel concentration in the inner and the intermediate layers than that of the outermost layer. In addition, a problem of lacking a reactor core reactivity at an initial stage is solved by disposing both of the fuel pellets together, even if burnable poisons of high concentration are used. This is because the extent of the lowering of the reactivity due to the burnable poison-incorporated fuels is mainly determined by the surface area thereof and the remaining period of the burnable poison is mainly determined by the concentration thereof. As a result, the burnup degree can be improved without lowering the reactor reactivity so much. (N.H.)

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

    International Nuclear Information System (INIS)

    Ragusa, Jean; Vierow, Karen

    2011-01-01

    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.

  19. 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.

  20. 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.

  1. Fuel cycle and waste management. 2. Design of a BWR Core with Over-moderated MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Francois, J.L.; Del Campo, C. Martin

    2001-01-01

    The use of uranium-plutonium mixed-oxide (MOX) fuel in light water reactors is a current practice in several countries. Generally one-third of the reactor core is loaded with MOX fuel assemblies, and the other two-thirds is loaded with uranium assemblies. Nevertheless, the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this work, the design of a boiling water reactor (BWR) core fully loaded with over-moderated MOX fuel designs was investigated. In previous work, the design of over-moderated BWR MOX fuel assemblies based on a 10 x 10 lattice was presented; these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. To increase the moderator-to-fuel ratio (MFR), two approaches were followed. In the first approach, 8 or 12 fuel rods were replaced by water rods in the 10x10 assembly, which increased the MFR from 1.9 to 2.2 and 2.4, respectively. These designs are called MOX-8WR and MOX-12WR, respectively, in this paper. In the second approach, an 11 x 11 lattice with 24 water rods (11 x 11-24WR) was designed, which is a design with a number of active fuel rods (88) very close to the standard MOX assembly (91). The fuel rod diameter is smaller to preserve the assembly dimensions, and in this last case, the MFR is 2.4. The calculations were performed with the CM-PRESTO three-dimensional steady-state simulator. The nuclear data banks were generated with the HELIOS system, and they were processed by TABGEN to produce tables of nuclear cross sections depending on burnup, void, and exposure weighted void (void history), which are used by CM-PRESTO. One base reload pattern was designed for a BWR/5 rated at 1931 MW(thermal), to be used with the different over-moderated assembly designs. The reload pattern has 112 fresh fuel assemblies (FFAs) out of a total of 444 fuel assemblies and was simulated during 20 cycles with the Haling strategy, until an equilibrium cycle of

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. Design and performance verification of fuel assembly and steam generator simulators for SMART reactor

    International Nuclear Information System (INIS)

    Ko, Y.-J.; Chu, I.-C.; Youn, Y.J.; Cho, Y.I.; Euh, D.J.

    2011-01-01

    The SMART reactor has been developed at KAERI, for the generation of electric power and also for seawater desalination. In order to verify the performance of the SMART design with respect to flow and pressure distribution, an experimental test facility named SCOP has been developed. For the purpose of preserving the flow distribution characteristics, SCOP is linearly reduced with a scaling ratio of 1/5. A CFD analysis was carried out to draw basic design parameters of the venturi tube and the perforated plates in a fuel assembly simulator. A CALIP, which is a flow and pressure drop calibration test facility, has been constructed to evaluate the pressure drop characteristic of fuel assembly and steam generator simulators. This paper shows the results of the actual performance verification and evaluation of fuel assembly and steam generator simulator, were evaluated using a CALIP. (author)

  4. Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

    International Nuclear Information System (INIS)

    Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

    2013-01-01

    Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

  5. Fuel assembly cleaning device

    International Nuclear Information System (INIS)

    Kikuchi, Akira.

    1981-01-01

    Purpose: To enable efficient and sufficient cleaning of a fuel assembly even in corners without disassembling the assembly and to effectively remove crud. Constitution: Cleaning water mixed with abrasive is injected into a fuel assembly contained within a cleaning device body to remove crud adhering to the fuel assembly. Since a coolant passage from the opening of the bottom surface is of the fuel assembly to the opening of the top surface is utilized as the cleaning water passage at this, the crud can be removed by the abrasive in the water stream even from narrow gaps of the fuel assembly. (Aizawa, K.)

  6. BRET fuel assembly dismantling machine

    International Nuclear Information System (INIS)

    Titzler, P.A.; Bennett, K.L.; Kelley, R.S. Jr.; Stringer, J.L.

    1984-08-01

    An automated remote nuclear fuel assembly milling and dismantling machine has been designed, developed, and demonstrated at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The machine can be used to dismantle irradiated breeder fuel assemblies from the Fast Flux Test Facility prior to fuel reprocessing. It can be installed in an existing remotely operated shielded hot cell facility, the Fuels and Materials Examination Facility (FMEF), at the Hanford Site in Richland, Washington

  7. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  8. Design of a mixed-oxide fuel assembly to be assessed as a lead test assembly in a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez, H.; Alonso, G.

    2001-01-01

    The open and the close cycle are the two alternatives to pursue during power generation. The reprocessing is a mature process that now shows a more competitive economic aspect, making it more attractive than ever. Mexico has not decided what to do with the existing and future depleted fuel assemblies that will be generated from the power operation, thus the direct disposal and the reprocessing are still being considered. To have enough arguments in one or the other alternatives it is necessary to make an assessment of both. This investigation focus in the MOX fuel design assuming that the reprocessing is the option to follow and looking for the lowest impact in power generation. The first step in a reprocessing program is to analyze the performance of four lead test assemblies (LTA's), thus in this investigation we design the corresponding MOX to be used as LTA's and assess their performance through one operational cycle. (author)

  9. Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1999-01-01

    The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

  10. 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.

  11. High utilization fuel assembly

    International Nuclear Information System (INIS)

    Camden, T.M. Jr.

    1986-01-01

    A nuclear fuel assembly is described comprising an array of parallel arranged guide tubes, an inlet nozzle attached to one end of the guide tubes, an outlet nozzle attached to the other end of the guide tubes, grids having the openings therethrough attached to and spaced along the length of the guide tubes, and of parallel arranged fuel rod assemblies each having an upper end and a lower end. The fuel rod assemblies are fitted within the openings in the grids, the fuel rod assemblies being arranged axially offset relative to each adjacent fuel rod assembly and comprising an upper fuel rod and a lower axially aligned fuel rod with a gap therebetween. The gap between the fuel rods each is axially offset relative to each adjacent gap so as to eliminate an axial gap across the core

  12. A SCWR core design with a conceptual fuel assembly using a cruciform moderator

    International Nuclear Information System (INIS)

    Bae, Kang Mok; Joo, Hyung Kook; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

    2005-01-01

    A super critical water cooled reactor (SCWR) system has a potential to compete with the advanced fossil plant by achieving a high thermal efficiency up to 44% and a plant simplification by eliminating steam generators, steam dryers, steam separators, and recirculation pumps. Due to these advantages, a SCWR is considered as one of the most promising nuclear plants for the Generation-IV (Gen-IV) system. As a first step of a feasibility study a rectangular fuel assembly with a cruciform solid moderator was suggested as a conceptual assembly design at the Korea Atomic Energy Research Institute (KAERI) for the SCWR on a thermal neutron spectrum. In this paper, based on the system parameters proposed by the Gen-IV road map, a preliminary SCWR core design was performed using a conceptual assembly design focused on the power shape control, reactivity coefficients, and cladding temperature limit

  13. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  14. Quality assurance in the procurement, design and manufacture of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    1983-01-01

    This Safety Guide provides requirements and recommendations for quality assurance programmes that are relevant for the unique features of the procurement, design, manufacture, inspection, testing, packaging, shipping, storage, and receiving inspection of fuel assemblies for nuclear power plants. The generic quality assurance requirements of the Code and related Safety Guides are referred to where applicable, and are duplicated in this document where increased emphasis is desirable

  15. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    International Nuclear Information System (INIS)

    Akiyama, Mamoru; Inoue, Akira; Miyazaki, Keiji; Abeta, Sadaaki; Hori, Keiichi; Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki.

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data

  16. Articulate fuel assembly

    International Nuclear Information System (INIS)

    Noyes, R.C.

    1978-01-01

    An articulated fuel assembly for the core of a fast spectrum reactor comprising an elongated shroud enclosing a cluster of fuel pins, a support foot assembly supporting the fuel assembly in the reactor core and an articulating connector link joining the support foot assembly and the lower end of the elongated shroud is described. The upper end of the elongated shroud and the support foot assembly are adapted to be fixedly restrained against lateral movement when the assembly is placed in the reactor core. The articulating connector link is such as to permit free lateral deflection of the lower end of the shroud relative to the upper end of the shroud and the foot assembly. Such an arrangement icreases the reliability of the fuel assembly and safely accommodates the physical distortions in the fuel assemblies caused by neutron induced swelling of the members and thermally induced expansions thereof by reducing stresses in the structural parts of the assembly and by insuring a negative reactivity for the core as the lower ends of the fuel assemblies are laterally displaced. 4 claims, 4 figures

  17. Fuel assemblies mechanical behaviour improvements based on design changes and loading patterns computational analyses

    International Nuclear Information System (INIS)

    Marin, J.; Aullo, M.; Gutierrez, E.

    2001-01-01

    In the past few years, incomplete RCCA insertion events (IRI) have been taking place at some nuclear plants. Large guide thimble distortion caused by high compressive loads together with the irradiation induced material creep and growth, is considered as the primary cause of those events. This disturbing phenomenon is worsened when some fuel assemblies are deformed to the extent that they push the neighbouring fuel assemblies and the distortion is transmitted along the core. In order to better understand this mechanism, ENUSA has developed a methodology based on finite element core simulation to enable assessments on the propensity of a given core loading pattern to propagate the distortion along the core. At the same time, the core loading pattern could be decided interacting with nuclear design to obtain the optimum response under both, nuclear and mechanical point of views, with the objective of progressively attenuating the core distortion. (author)

  18. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

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

  19. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

    2009-01-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  20. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1980-01-01

    A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

  1. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  2. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Helmersson, S.

    1982-05-01

    The fuel assembly has a square-shaped cross section and it is put together of four quadratic assemblies each having seventeen positions for fuel rods, which are situated in a lattice formed by a pattern of triangles and squares. Nine of the positions correspond to the junction of a square lattice which has four squares, whereas eight rods are outside the quadratic past. (G.B.)

  3. Application of the inverse generalized perturbation theory to the optimization of fuel assembly design

    International Nuclear Information System (INIS)

    Dall’Osso, Aldo

    2016-01-01

    Highlights: • The method is useful to determine optimal parameters in multi-objective problems. • The perturbation equations are exact and need only one calculation of the adjoints. • Examples as the determination of the water density maximizing k-eff are presented. - Abstract: Many methods have been proposed to optimize fuel assembly design, most of them based on metaheuristic techniques. The method presented here is based on the inverse perturbation theory. Parameters to be optimized are some isotope densities, such as Gd, 235 U, 239 Pu. The optimization is constrained to some target values of relevant reactor observables, such as the breeding ratio, the reactivity loss rate during the fuel cycle, the maximum reactivity, the spectral index. The method is fit to solve multi-objective problems. In the examples, 2 or 3 simultaneous objectives are determined.

  4. Transfer of fuel assemblies

    International Nuclear Information System (INIS)

    Vuckovich, M.; Burkett, J. P.; Sallustio, J.

    1984-01-01

    Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly

  5. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    The nuclear fuel assembly described includes a cluster of fuel elements supported at a distance from each other so that their axes are parallel in order to establish secondary channels between them reserved for the coolant. Several ducts for an auxiliary cooling fluid are arranged in the cluster. The wall of each duct is pierced with coolant ejection holes which are placed circumferentially to a pre-determined pattern established according to the position of the duct in the cluster and by the axial distance of the ejection hole along the duct. This assembly is intended for reactors cooled by light or heavy water [fr

  6. 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.

  7. Final report: Seven-layer membrane electrode assembly - an innovative approach to PEM fuel cell design

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, A.

    2005-07-01

    Costs of materials and fabrication, rather than appropriateness of technology, are the major barriers to the sales of fuel cells. With the objective of reducing costs, potential alternative component materials for (a) the fluid flow plate (FFP) and (b) the gas diffusion layers were investigated. The concept of a 7-layer membrane electrode assembly (MEA), in which components are bonded into a unitised module, was also studied. The advantages of the bonded cell, and the flow field design, are expounded. Low-cost carbon particle composites were developed for the FFPs. The modular 7-layer MEA has an order of magnitude saving over current materials. Overall, the study has led to a greater volumetric power output, lower costs and greater reliability. The work was carried out by Morgan Group Technology Limited and funded by the DTI.

  8. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

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

  9. 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

  10. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

  11. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Takeda, Tadashi; Sato, Kenji; Goto, Masakazu.

    1984-01-01

    Purpose: To facilitate identification of a fuel assembly upon fuel exchange in BWR type reactors. Constitution: Fluorescent material is coated or metal plating is applied to the impressed portion of a upper tie plate handle of a fuel assembly, and the fluorescent material or the metal plating surface is covered with a protective membrane made of transparent material. This enables to distinguish the impressed surface from a distant place and chemical reaction between the impressed surface and the reactor water can be prevented. Furthermore, since the protective membrane is formed such that it protrudes toward the upper side relative to the impressed surface, there is no risk of depositions of claddings thereover. (Moriyama, K.)

  12. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  13. Reactor fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1980-01-01

    A nuclear reactor fuel assembly having a lower end fitting and actuating means interacting therewith for holding the assembly down on the core support stand against the upward flow of coolant. Locking means for interacting with projections on the support stand are carried by the lower end fitting and are actuated by the movement of an actuating rod operated from above the top of the assembly. In one embodiment of the invention the downward movement of the actuating rod forces a latched spring to move outward into locking engagement with a shoulder on the support stand projections. In another embodiment, the actuating rod is rotated to effect the locking between the end fitting and the projection. (author)

  14. Instrumented fuel assembly for Dhruva [Paper No.:P3

    International Nuclear Information System (INIS)

    Pappu, Ajit; Limaye, S.P.; Singh, D.B.

    1993-01-01

    Dhruva is a heavy water cooled and moderated research reactor and uses natural uranium metal as fuel. The fuel is cladded with aluminum. Dhruva fuel design was carried out based on the estimated maximum fuel clad temperature. In order to validate the design, instrumented fuel assembly is developed. This assembly consists temperature sensors for in core temperature measurement. This paper describes details of instrumented fuel assembly and results of in pile experiments conducted using the assembly. (author). 5 figs., 2 tabs

  15. Casette for storage of spent fuel assemblies

    International Nuclear Information System (INIS)

    Ericsson, S.

    1992-01-01

    Describes a design of a casette for spent fuel storage in a fuelstorage pool. The new design, based on flexible spacers, allows the fuel assemblies to be packed more compact and the fuel storage pool used in a more economic way

  16. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

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

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo

    2012-01-01

    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.)

  18. 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.)

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hoshi, Masaya; Makihara, Yoshiaki.

    1985-01-01

    Purpose: To limit a bypass flow by inhibiting or restricting the lateral flow of coolants between lower nozzle legs of a nuclear fuel assembly, so that the flow speed of a jet stream flowing through the gaps between buffle plates into the reactor core is not increased. Constitution: The lower nozzle of a fuel assembly comprises an upper plate, an enclosure and legs, in which flow apertures are perforated in the enclosure, the area for the flow apertures and the slit are set to less than predetermined values, and the flow apertures are arranged so that they are situated within the gaps between the lower end of the buffle plate and the lower reactor core plate. As the result, since the jet stream from the gaps between the buffle plates can be so decreased as the effect thereof on the fuel rods is negligible, measurement for the size of the gap between the buffle plates upon periodical inspection is no more necessary, thereby enabling to shorten the time of the periodical inspection and reduce the exposure dose. (Kamimura, M.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  1. 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.

  2. Fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Saito, Shozo; Kawahara, Akira.

    1975-01-01

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

  3. Assembling method for nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Kamiwaki, Yoshiharu; Ono, Shunji.

    1996-01-01

    Control rod guide tubes are inserted and assembled in to predetermined openings of a plurality of support lattices arranged in rows at predetermined distance. Sheath heaters are inserted to the guide tubes. The sheath heater comprises a plurality of heater elements therein which are connected to a temperature controller. The temperature of each portion of the sheath heater is controlled by the temperature controller. When the temperature of each portion of the guide tube is made uniform, MOX fuel rods are inserted to vacant openings of the support lattices. This changes the circumferential temperature of the guide tube, but the heat generation amount of each heater element is controlled suitably by the temperature controller. Accordingly, even if MOX fuel rods are inserted as a heat generation member, fuel assembly can be assembled with no assembling errors. (I.N.)

  4. Reactor fuel assembly fastening

    International Nuclear Information System (INIS)

    Formanek, F.J.; Schukei, G.E.

    1980-01-01

    A nuclear fuel assembly is described, adapted to be locked into first mating surfaces on a core support stand, comprising a lower end fitting having posts for resting on the stand; elongated hook members pivotally connected at one end to the lower end fitting and having a second mating surface at the other end to engage the first mating surfaces; actuating means located between the posts on the lower end fitting and being vertically movable relative to the end fitting; and rigid links pivotally attached at one end to the hook members intermediate the connection of the hook members to the end fitting and the second mating surface and pivotally attached at the other end to the actuating means, the link having a length between the pivoted connections such that the second mating surface on the hook members locks into engagement with the first mating surfaces on the stand as the links approach the horizontal. (author)

  5. Nuclear fuel: modelling the advanced plutonium assembly

    International Nuclear Information System (INIS)

    Kaoua, Th.; Lenain, R.

    2004-01-01

    The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

  6. Experience in WWER fuel assemblies vibration analysis

    International Nuclear Information System (INIS)

    Ovtcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.

    2003-01-01

    It is stated that the vibration studies of internals and the fuel assemblies should be conducted during the reactor designing, commissioning and commercial operation stages and the analysis methods being used should complement each other. The present paper describes the methods and main results of the vibration noise studies of internals and the fuel assemblies of the operating NPPs with WWER reactors, as an example of the implementation of the comprehensive approach to the analysis on equipment flow-induced vibration. At that, the characteristics of internals and fuel assemblies vibration loading were dealt jointly as they are elements of the same compound oscillating system and their vibrations have the interrelated nature

  7. Fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    Leonard, B.H. Jr.

    1975-01-01

    A description is given of a fuel assembly for a nuclear reactor comprising a plurality of elongated plate-like fuel bearing elements of the same length and width, paired longer than they are wide and assembly spacer members having means defining opposed spaced notches for receiving the side edges of said elongated plate-like fuel bearing elements, and means for securing said plate-like fuel bearing elements to said paired assembly spacer members with the side edges of said plate-like elements engaged in opposite notches in said paired assembly spacer elements so as to secure said fuel bearing elements in side by side spaced relation in a staggered arrangement transversely so as to conform to a diamond shaped profile in which opposite sides are parallel and opposite angles are substantially 60 0 and substantially 120 0

  8. Nuclear reactor fuel assembly spacer grids

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    Designs of nuclear reactor fuel assembly spacer grids for supporting and spacing fuel elements are described which do not utilize resilient grid plate protrusions in the peripheral band but retain the advantages inherent in the combination resilient and rigid protrusion cells. (U.K.)

  9. Fuel rod and fuel assembly

    International Nuclear Information System (INIS)

    Takekawa, Tetsuya.

    1993-01-01

    Burnable poisons are contained in a portion of a pellet constituting a fuel rod. A distribution density of the burnable poison-containing pellets and a concentration of the burnable poisons in the pellet are varied depending on the axial position of the fuel rod. That is, the distribution density of the burnable poison containing-pellets is increased at the central portion of the fuel rod and it is decreased at both ends thereof, and a concentration of the burnable poisons of the burnable poison containing-pellet disposed at the end portions thereof is decreased to less than a concentration of the burnable poison-containing pellet at the central portion. With such a constitution, a central peaking at an early stage of the combustion cycle is decreased. Accordingly, power at the central portion is increased than that in the end portions at the latter half of the cycle, to flatten the power distribution. Further, a burnable poison concentration of the pellets at the end portions is decreased to promote burning of burnable poisons at the end portions which are less burnable relatively, thereby enabling to prevent worsening of neutron economy. (T.M.)

  10. Burnable absorber-integrated Guide Thimble (BigT) - 1. Design concepts and neutronic characterization on the fuel assembly benchmarks

    International Nuclear Information System (INIS)

    Yahya, Mohd-Syukri; Yu, Hwanyeal; Kim, Yonghee

    2016-01-01

    This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named 'Burnable absorber-integrated Guide Thimble' (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal-hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17x17 and 16x16 fuel assembly lattices. For the 17x17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16x16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library. (author)

  11. A classification scheme for LWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  12. International LWR fuel design

    International Nuclear Information System (INIS)

    Bjornard, T.A.

    1990-01-01

    This paper reports on current Light Water Reactor (LWR) fuel designs throughout the world have many basic features in common, such as, cylindrical UO 2 fuel pellets contained in zirconium alloy cladding, helium fill gas inside the fuel rod, and a skeleton structure consisting of end fittings and spacers. These, as a minimum, are features one would find in virtually every fuel design for every LWR. To the eye of the uninitiated, one design would probably appear to be much like any other: i.e., rods of roughly equal diameter and length held together by a skeleton structure. But--nuclear fuel assembly design is a business where a few mils, a few ppm, a few degrees in temperature, or a fraction of a percent in a key parameter make all the difference between a merely acceptable design and a superior design

  13. FUEL ROD ASSEMBLY

    Science.gov (United States)

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  14. Establishment of China Nuclear Fuel Assembly Database

    International Nuclear Information System (INIS)

    Chen Peng; Jin Yongli; Zhang Yingchao; Lu Huaquan; Chen Jianxin

    2009-01-01

    China Nuclear Fuel Assembly Database (CNFAD) is developed based on Oracle system. It contains the information of fuel assemblies in the stages of its design, fabrication and post irradiation (PIE). The structure of Browser Sever is adopted in the development of the software, which supports the HTTP protocol. It uses Java interface to transfer the codes from server to clients and make the sources of server and clients be utilized reasonably and sufficiently, so it can perform complicated tasks. Data in various stages of the fuel assemblies in Pressure Water Reactor (PWR), such as the design,fabrication, operation, and post irradiation examination, can be stored in this database. Data can be shared by multi users and communicated within long distances. By using CNFAD, the problem of decentralization of fuel data in China nuclear power plants will be solved. (authors)

  15. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

    Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

  16. 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

  17. Metallic fuel design development

    International Nuclear Information System (INIS)

    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 design technology on metallic fuel

  18. Fuel assembly insertion system

    International Nuclear Information System (INIS)

    Barkhurst, D.J.

    1987-01-01

    This patent describes a nuclear reactor facility having fuel bundles: a system for the insertion of a fuel bundle into a position where vertically arranged fuel bundles surround and are adjacent the system comprising, in combination, separate and individual centering devices secured to and disposed on top of each fuel bundle adjacent the position. Each such centering device has a generally box-like cap configuration on the upper end of each fuel bundle and includes: a top wall; first and second side walls, each secured along and upper edge to the top wall; a rear plate attached along opposite vertical edges to the first and second side walls; a front inclined wall joined along an upper edge to the top to the wall and attached along opposite vertical edges first and second side walls; pad means secured to the lower edge of the first and second side walls, the front inclined wall and the rear plate for mounting each centering device on top of an associated fuel bundle; pin means carried by at least two of the pad means engageable with an associated aperature for locating and laterally fixing each centering device on top of its respective fuel bundle. Each front inclined wall of each of the centering devices is orientated on top of its respective fuel bundle to slope upwardly and away from the position where upon downward insertion of a fuel bundle any contact between the lower end of the fuel bundle inserted with a front inclined wall of a centering device will laterally deflect the fuel bundle. Each centering device further includes a central socket means secured to the top wall, and an elongated handling pole pivotally attached to the socket

  19. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  20. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

    The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

  1. A drying system for spent fuel assemblies

    International Nuclear Information System (INIS)

    Suikki, M.; Warinowski, M.; Nieminen, J.

    2007-06-01

    The report presents a proposed drying apparatus for spent fuel assemblies. The apparatus is used for removing the moisture left in fuel assemblies during intermediate storage and transport. The apparatus shall be installed in connection with the fuel handling cell of an encapsulation plant. The report presents basic requirements for and implementation of the drying system, calculation of the drying process, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components have not been included. The report also describes actions for possible malfunction and fault conditions. An objective of the drying system for fuel assemblies is to remove moisture from the assemblies prior to placing the same in a disposal canister for spent nuclear fuel. Drying is performed as a vacuum drying process for vaporizing and draining the moisture present on the surface of the assemblies. The apparatus comprises two pieces of drying equipment. One of the chambers is equipped to take up Lo1-2 fuel assemblies and the other OL1-2 fuel assemblies. The chambers have an internal space sufficient to accommodate also OL3 fuel assemblies, but this requires replacing the internal chamber structure for laying down the assemblies to be dried. The drying chambers can be closed with hatches facing the fuel handling cell. Water vapour pumped out of the chamber is collected in a controlled manner, first by condensing with a heat exchanger and further by freezing in a cold trap. For reasons of safety, the exhaust air of vacuum pumps is further delivered into the ventilation outlet duct of a controlled area. The adequate drying result is ascertained by a low final pressure of about 100 Pa, as well as by a sufficient holding time. The chamber is built for making its cleaning as easy as possible in the event of a fuel rod breaking during a drying, loading or unloading

  2. Irradiated fuel assembly restoration equipment

    International Nuclear Information System (INIS)

    Guironnet, L.

    1993-01-01

    Analysis of in-plant fuel operating experience shows that assembly damage has a variety of causes: Handling incidents; external hazards during operation; wear and perforation; fuel manufacturing defects and other causes. Depending on the seriousness of the damage and the burnup level reached by assemblies, several repair possibilities arise: If they are leaktight and if grid distortion is minimal and does not jeopardize rod restraint, they are reloaded, subject to local servicing for restoring acceptable grid geometry. In all other cases, the assemblies have to be restored either by replacing the damaged skeleton or be replacing the leaking rod(s). This paper presents methods, equipments and the FRAGEMA experience in assemblies repair. (author). 5 pictures, 3 diagrams

  3. Assembly tool design

    International Nuclear Information System (INIS)

    Kanamori, Naokazu; Nakahira, Masataka; Ohkawa, Yoshinao; Tada, Eisuke; Seki, Masahiro

    1996-06-01

    The reactor core of the International Thermonuclear Experimental Reactor (ITER) is assembled with a number of large and asymmetric components within a tight tolerance in order to assure the structural integrity for various loads and to provide the tritium confinement. In addition, the assembly procedure should be compatible with remote operation since the core structures will be activated by 14-MeV neutrons once it starts operation and thus personal access will be prohibited. Accordingly, the assembly procedure and tool design are quite essential and should be designed from the beginning to facilitate remote operation. According to the ITER Design Task Agreement, the Japan Atomic Energy Research Institute (JAERI) has performed design study to develop the assembly procedures and associated tool design for the ITER tokamak assembly. This report describes outlines of the assembly tools and the remaining issues obtained in this design study. (author)

  4. Nuclear fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.; Butterfield, C.E.; Waite, E.

    1979-01-01

    A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

  5. FOCUS-type fuel assembly for PWRs

    International Nuclear Information System (INIS)

    Aisch, F.W.; Fuchs, H.P.; Lettau, H.

    1994-01-01

    In recent years the development efforts for Siemens PWR fuel assemblies were mainly concentrated on reducing the fuel cycle costs and increasing the operational reliability of the fuel assemblies. The first objective was aimed at increasing the average discharge burnup to >50 MWd/kgU and increasing the critical heat flux. The high envisaged burnup required to develop a corrosion resistant cladding tube outside the Zry-4 range. The decision was made to use a Duplex cladding tube consisting of a corrosion optimized outer layer on a Zry-4 base material. A ZrSnFeCr alloy with reduced tin content was chosen for the outer layer. The critical heat flux could be increased by introducing mixing vanes on the spacer grids within the active length. To reach the second objective, reliable avoidance of spacer grid damage during core loading and unloading and reduction of fuel rod defects by debris fretting, the spacer grid corners were improved and a debris separation grid was developed. These design improvements were introduced into the new FOCUS-type fuel assembly. The name FOCUS stands for 'Fuel assembly with Optimized Cladding and Upgraded Structure'. (orig.)

  6. Using FARIS [Fuel Assembly Repair and Inspection Station] for assembly clean-up and debris removal

    International Nuclear Information System (INIS)

    Tucker, J.S.; Sapyta, J.J.

    1990-01-01

    Because fuel inspection and repair tasks are commonly done on the critical path during plant refuelling outages, they must be completed quickly and efficiently with minimal costs. To fulfil these demands, the Babcock and Wilcox Fuel Company has designed a Fuel Assembly Repair and Inspection Station (FARIS) for fuel assembly clean-up and debris removal in Pressurized Water Reactors. The system is portable and can also be used for carrying out visual inspections on fuel assemblies, spacer grid repair, fuel rod oxide thickness measurements and for fuel rod water channel inspections. (author)

  7. Implementation of a quality assurance system for the design and manufacturing of fuel assembly MTR-plate type

    International Nuclear Information System (INIS)

    Koll, J.H.

    1987-01-01

    Since more than 30 years ago, fuel assemblies (FA) of the MTR-Plate type, for research reactors, have been developed and produced using well known technologies, with different methods for the design, manufacturing, quality control and subsequent verification of FA behaviour, as well as of the design data. The FA and its reliability has been improved through the recycling of the obtained information. No nuclear accidents or major incidents have taken place that can be blamed to FA due to design, manufacturing or its use. Since the 70's, the use of Quality Assurance methodology has been increased, especially for Nuclear Power Plants, in order to ensure safety for these reactors. The use of QA for reactors for research, testing or other uses, has also been steadily increased, not only due to safety reasons, but also because of its convenience for a good operation, being presently a common requirement of the operator of the installation. Herewith is described the way the QA system that has been developed for the design, manufacturing, quality control and supply of MTR-plate type FA, at the Development Section of the Argentine Atomic Energy Commission (CNEA). (Author)

  8. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    Science.gov (United States)

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  9. 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.

  10. Appearance detection device for fuel assembly

    International Nuclear Information System (INIS)

    Matsuoka, Toshihiro

    1998-01-01

    The prevent invention provides an appearance detection device which improves accuracy of images on a display and facilitates editing and selection of images upon detection of appearance of a reactor fuel assembly. Namely, the device of the present invention comprises (1) television cameras movable along fuel assemblies of a reactor, (2) a detection means for detecting the positions of the television cameras, (3) a convertor for converting analog image signals of the television cameras to digital image signals, (4) a memory means for sampling a predetermined portion of the images of the television camera and storing it together with the position signal obtained by the detection means and (5) a computer for selecting a plurality of images and positions from the above-mentioned means and joining them to one or a plurality of static images of the fuel assembly. At least two television cameras are disposed oppositely with each other. Then, position signals of the television cameras are designated by the stored sampling signals, and the fuel assembly at the position can be displayed quickly. It is scrolled, compressed or enlarged and formed into images. (I.S.)

  11. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. Liaison based assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Ames, A.; Kholwadwala, D.; Wilson, R.H.

    1996-12-01

    Liaison Based Assembly Design extends the current information infrastructure to support design in terms of kinematic relationships between parts, or liaisons. These liaisons capture information regarding contact, degrees-of-freedom constraints and containment relationships between parts in an assembly. The project involved defining a useful collection of liaison representations, investigating their properties, and providing for maximum use of the data in downstream applications. We tested our ideas by implementing a prototype system involving extensions to Pro/Engineer and the Archimedes assembly planner. With an expanded product model, the design system is more able to capture design intent. When a product update is attempted, increased knowledge availability improves our ability to understand the effect of design changes. Manufacturing and analysis disciplines benefit from having liaison information available, so less time is wasted arguing over incomplete design specifications and our enterprise can be more completely integrated.

  13. Determination of the radioactive inventory of a fuel assembly from a U3O8 design core using ORIGEN 2.1 code

    International Nuclear Information System (INIS)

    Castro, Jose; Ticona, Braulio; Madariaga, Marcelo

    2014-01-01

    This paper shows a methodology to determine the radioactive inventory of a fuel assembly of the RP-10 design core, which was proposed in 1988, using the ORIGEN 2.1 code, which allows to determine the activity of the 52 most characteristic fission products, its growth in activity during reactor operation under the terms of the design and evolution of decay of the fission products after 4 hours after the reactor shutdown, which conservatively, a fuel element represents an average fraction of the considered power in the radioactive inventory assessment. (authors).

  14. Design optimization of a T mixing vane in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Jung, Sang-Ho; Moon, Mi-Ae; Kim, Kwang-Yong

    2009-01-01

    The purposes of present work are to analyze the convective heat transfer with three-dimensional Reynolds-averaged Navier-Stokes analysis, and to optimize shape of the mixing vane using the analysis results. PLUS7 that is designed by KNF and Westinghouse is used as reference geometry. Shear stress transport turbulence model is used as a turbulence closure. Two bend angles of mixing vane are selected as design variable. The objective function is defined as a combination of inverse of heat transfer rate and friction loss. Response surface method is employed as an optimization technique. The calculation domains of 1x2 geometry are analyzed with translational and rotational periodic boundary conditions which take flow directions into account. The fluid flow and heat transfer characteristics have been explained through velocity vectors, streamlines and Nusselt numbers. The results show that the optimized geometry improves the heat transfer performance of the mixing vane with a relatively small pressure drop increment and has higher Critical Heat Flux. (author)

  15. Core fuel management using TVS-2M fuel assembly and economic analysis

    International Nuclear Information System (INIS)

    Xu Min; Wang Hongxia; Li Youyi

    2014-01-01

    To improve the economic efficiency, TVS-2M fuel assembly was considered to apply in Tianwan Nuclear Power Plant units 3, 4. Using KASKAD program package, a preliminary research and design was carried out for the Tianwan Nuclear Power Plant loading TVS-2M fuel assembly from the first cycle to equilibrium cycle. An improved fuel management program was obtained, and the economic analysis of the two fuel management programs with or without TVS-2M assembly was studied. The analysis results show that TVS-2M fuel assembly can improve the economic efficiency of the plant remarkably. (authors)

  16. Simulated nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Berta, V.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

  17. 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.

  18. Mission MOX Fuel Physics Design--Preliminary Equilibrium MOX Assembly Design and Expected Operating Power for Existing Balakovo Fuel Management Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-09-28

    Among various versions of excess weapons-grade plutonium handling the most preferred in Russia is its burning in power reactors. This is accounted for by the desire to utilize the power value of weapons-grade plutonium and the potentialities of the existing nuclear industry complex. In Russia the versions of burning weapons-grade plutonium in the VVER-, BN-, and HTGR-type power reactors are being developed. However the analysis of the current structure of nuclear power and the energy strategy reveals that in the coming years the VVER-1000-type (designs B-320 and B-392) as well as the VVER-640 reactor (design B-407) now under development appear to be the most promising for this purpose. The experience with the use of mixed uranium/plutonium fuel in the LWR, gained in the West and the preliminary studies carried out in Russia show that weapons-grade plutonium may be actually used as fuel for the Russian VVER reactors. At present Russia has 7 operating VVER-1000 of total installed capacity 7 GWe, 11 reactors of this type are in operation in Ukraine, and 2 in Bulgaria. Before 2003 it is planned to put into operation 2 VVER-1000 units more in Russian and at least 2 units in Ukraine.

  19. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  20. Magnetic scanning of LWR fuel assemblies

    International Nuclear Information System (INIS)

    Fiarman, S.; Moodenbaugh, A.

    1980-01-01

    Nondestructive assay (NDA) techniques are available both for fresh and spent fuel, but generally are too time consuming and do not uniquely identify an assembly. A new method is reported to obtain a signature from a magnetic scan of each assembly. This scan is an NDA technique that detects magnetic inclusions. It is potentially fast (5 min/assembly), and may provide a unique signature from the magnetic properties of each fuel assembly

  1. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.

    1982-01-01

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

  2. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    Moffett, S.D.

    1985-01-01

    This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate

  3. Development of the new generation fuel elements and assemblies for research reactors

    International Nuclear Information System (INIS)

    Solonin, M.I.; Vatulin, A.V.; Stetskij, Yu.A.; Dobrikova, I.V.; Ivanov, A.V.; Kruglov, A.A.; Leonov, E.G.

    2002-01-01

    The fuel elements and fuel assemblies for the research reactors (RR) of the new generation are described. In particular, the designs of the rod and tubular fuel elements and fuel assemblies are developed and their optimal dimensions are determined in three versions. These fuel elements make it possible to arrange fuel assemblies for any RR without changing the dimensions and configuration of the existing RR cores. The technology for the fuel elements manufacture is worked out under laboratory conditions [ru

  4. Nuclear fuel assembly seismic amplitude limiter

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1977-01-01

    The ability of a nuclear reactor to withstand high seismic loading is enhanced by including, on each fuel assembly, at least one seismic grid which reduces the magnitude of the possible lateral deflection of the individual fuel elements and the entire fuel assembly. The reduction in possible deflection minimizes the possibility of impact of the spacer grids of one fuel assembly on those of an adjacent fuel assembly and reduces the magnitude of forces associated with any such impact thereby minimizing the possibility of fuel assembly damage as a result of high seismic loading. The seismic grid is mounted from the fuel assembly guide tubes, has greater external dimensions when compared to the fuel assembly spacer grids and normally does not support or otherwise contact the fuel elements. The reduction in possible deflection is achieved through reduction of the clearance between adjacent fuel assemblies made possible by the use in the seismic grid of a high strength material characterized by favorable thermal expansion characteristics and minimal irradiation induced expansion

  5. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Mochida, Takaaki.

    1987-01-01

    Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

  6. Experimental study of new generation WWER-1000 fuel assemblies at JSC NCCP

    International Nuclear Information System (INIS)

    Enin, A.; Rozhkov, V.; Sinikov, Y.; Ustimenko, A.; Shustov, M.

    2003-01-01

    An experimental program for the study of fuel assembly thermomechanical stability has been established together with RF SSC IPPE and Russian Scientific Center Kurchatov Institute. Assembly fragments and small dummy models of fuel assembly skeletons and fuel rod bundles have been used for the tests. The test results are used for the design selection, verification of the design codes and substantiation of operating capacity of fuel assemblies with a rigid skeleton. The mechanical characteristics of units make it possible to perform fuel assembly strength and rigidity calculations, including the cases of abnormal operation. The mechanical characteristics of the skeleton and fuel rod bundle dummy models make it possible to check for the adequacy of the fuel assembly design model. The mechanical characteristics obtained during fuel rods bundle push through experiments make it possible to substantiate the fuel assembly serviceability under the conditions of fuel rods bundle and skeleton interaction

  7. Criticality safety evaluation report for FFTF 42% fuel assemblies

    International Nuclear Information System (INIS)

    Richard, R.F.

    1997-01-01

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC)

  8. On-site scanning and operation equipment for fuel assemblies and fuel pencils

    International Nuclear Information System (INIS)

    Lavoine, O.; Leseur, A.

    1984-10-01

    Presentation of equipments utilizable in cooling pools and which meet the two following objectives: - ensure on-site visual and measured control allowing fast evaluation of the mechanical behaviour of the fuel assemblies and fuel pencils, - extract pencils from assemblies designed for this purpose, so they can be transfered to hot cells where their mechanical, chemical and physical characteristics may be observed. After a review of the various devices used in France for this end (in service or under construction), design, operation and results are presented for three of them (fuel assemblies and fuel pencils control apparatus and removable pencil handling equipment) [fr

  9. Locking support for nuclear fuel assemblies

    Science.gov (United States)

    Ledin, Eric

    1980-01-01

    A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

  10. Nuclear reactor seismic fuel assembly grid

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1977-01-01

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

  11. Reconstitutable fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1981-01-01

    A reconstitutable fuel assembly for a nuclear reactor with a mechanical arrangement for connecting control rod guide thimbles to the top and bottom nozzle plates of a fuel assembly. Sleeves enclosing control rod guide thimbles interconnect the top and bottom nozzle plates and the fuel assembly upper and lower spacer grid. Each sleeve is secured to the respective nozzle plate by retaining rings disposed on opposite sides. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms. (author)

  12. Impact analysis of spent fuel jacket assemblies

    International Nuclear Information System (INIS)

    Aramayo, G.A.

    1994-01-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered

  13. The single SNR fuel assembly container (ESBB) to transport unirradiated SNR 300 fuel assemblies

    International Nuclear Information System (INIS)

    Hilbert, F.; Hottenrott, G.

    1998-01-01

    In this paper a new type B(U) package design is presented. The Single SNR Fuel Assembly Container (ESBB) is designed for the transport and storage of a single SNR 300 fuel assembly. This package is the main component for the future interim storage of the fuel assemblies in heavy storage casks. Its benefits are that it is compatible with the Category I transport system of Nuclear Cargo + Service NCS) used in Germany and that it can be easily handled at the current storage locations as well as in an interim storage facility. In total 205 fuel assemblies are currently stored in Hanau, Germany and Dounreay, U.K. Former studies have shown, that heavy transport and storage casks can be handled there only with considerable efforts. But the required category I transport to an interim storage is not reasonably feasible. To overcome these problems the ESBB was designed. It consists of a stainless steel tube with welded bottom, a welded plug as closure system and shock absorbers 26 packages at maximum can be transported in one batch with the NCS security vehicle. The safety analysis shows that the package complies with IAEA 1996. Standard calculations methods and computer codes like HEATING 7.2 (Childs 1993) have been used for the analysis. Criticality safety assessment is based on conservative assumptions as required in IAEA 1996. Drop tests carried out by BAM will be used to verify the design. These tests are scheduled for mid 1998. For the validation of the design prototypes have already been manufactured. Handling tests show that the design complies with the requirements. Preliminary drop tests show that the certification drop tests will be passed positively. (authors)

  14. Framatome experience in fuel assembly repair and reconstitution

    International Nuclear Information System (INIS)

    Leroy, G.

    1998-01-01

    Since 1985, FRAMATOME has build up extensive experience in the poolside replacement of fuel rods for repair or R and D purposes and the reconstitution of fuel assemblies (i.e. replacement of a damaged structure to enable reuse of the fuel rod bundle). This experience feedback enables FRAMATOME to improve in steps the technical process and the equipment used for the above operations in order to enhance their performance in terms of setup, flexibility, operating time and safety. In parallel, the fuel assembly and fuel rod designs have been modified to meet the same goals. The paper will describe: - the overall experience of FRAMATOME with UO 2 fuel as well as MOX fuel; the usual technical process used for fuel replacement and the corresponding equipment set; - the usual technical process for fuel assembly reconstitution and the corresponding equipment set. This process is rather unique since it takes profit of the specific FRAMATOME fuel assembly design with removable top and bottom nozzles, so that fuel rods insertion by pulling through in the new structure is similar to what is done in the manufacturing plant; - the usual inspections done on the fuel rods and/or the fuel assembly; - the design of the new reconstitution equipment (STAR) compared with the previous one as well as their comparative performance. The final section will be a description of the alternative reconstitution process and equipment used by FRAMATOME in reactors in which the process cannot be used for several reasons such as compatibility or administrative authorization. This process involves the pushing of fuel rods into the new structure, requiring further precautions. (author)

  15. Optical matrix for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Romero G, M.; Gonzaga O, A.

    1996-01-01

    In order to detect the presence of fuel rods, it was selected a reflection optical transducer, which provides a measurable electrical signal when the beam at a certain distance is interrupted then there is a reflection causing a excitation to the sensor that provides a change of state at the output of transducer. This step is coupled through an operational amplifier which drives the opto coupler circuit isolating this step of the interface and a personal computer. This work presents the description of components, designs, signal coupler and opto isolater circuit, interface circuit and tutorial assemble program. (Author)

  16. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Schluderberg, D.C.

    1981-01-01

    A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

  17. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2009-01-01

    A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

  18. Irradiated MTR fuel assemblies sipping test

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab.

  19. Irradiated MTR fuel assemblies sipping test

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A.

    1997-01-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab

  20. Benches for fuel assembly inspection and repair in the USA

    International Nuclear Information System (INIS)

    Rabchun, A.V.

    1988-01-01

    Design of two benches for inspecting and repairing fuel assemblies developed by Westinghouse and ANF firms is described. Main types of repair and a system for damaged fuel element diagnostics are considered and examples of the bench trial operation are presented

  1. TracWorksTM-global fuel assembly data management

    International Nuclear Information System (INIS)

    Cooney, B.F.

    1997-01-01

    The TracWorks TM Data Management System is a workstation-based software product that provides a utility with a single, broadly available, regularly updated source for virtually every data item available for a fuel assembly or core component. TracWorks is designed to collect, maintain and provide information about assembly and component locations and movements during the refueling process and operation, assembly burnup and isotopic inventory (both in-core and out-of-core), pin burnup and isotopics for pins that have been removed from their original assemblies, assembly and component inspection results (including video) and manufacturing data provided by the fabrication plant

  2. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  3. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Hutchinson, J.J.

    1978-01-01

    A nuclear reactor fuel assembly in which the end fittings may be easily removed after the assembly has been irradiated so that defective fuel rods may be replaced or special fuel or burnable poison rods inserted therein. The fuel assembly is of the type wherein structural support is provided by several vertically extending hollow structural members attached at opposite ends to upper and lower end fittings. The upper and lower end fittings each comprise an end plate and means extending therefrom for alignment and support of the assembly within the reactor core. Threaded joints between the hollow structural members and the means for alignment form the connections between the hollow structural members and the upper and lower end fittings

  4. 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.

  5. Nuclear fuel assembly and process

    International Nuclear Information System (INIS)

    Grubb, W.T.

    1978-01-01

    Rupture of boiling water reactor nuclear fuel cladding resulting from embrittlement caused by fission product cadmium is prevented by adding the stoichiometrically equivalent amount of gold, silver or palladium to the fuel

  6. Overview of fuel assembly structure analysis in regard to reactor safety

    International Nuclear Information System (INIS)

    Kim, J.D.

    1981-01-01

    In the reactor technique many special subjects are involved for the fuel assembly analysis especially; neutron physics and reactor chemistry, thermalhydraulics, laboratories, structural mechanics and fuel assembly manufacture. In this variety of fuel assembly analysis, design and manufacture the quality assurance is very essential on the interfaces. An overview of the interfaces is necessary besides of the special knowledge on the own subject. This paper deals with fuel assembly interrelation, design requirements and with a calculation model for the dynamic investigation of fuel assembly structure for the vertical direction. (author)

  7. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.

    1981-01-01

    An improved fuel sub-assembly for a liquid metal cooled fast breeder reactor, is described, in which fatigue damage due to buffeting by cross-current flows is reduced and protection is provided against damage by contact with other reactor structures during loading and unloading of the sub-assembly. (U.K.)

  8. Research of core fuel management using TVS-2M fuel assemblies in VVER

    International Nuclear Information System (INIS)

    Wang Hongxia; Xu Min

    2014-01-01

    Using KASKAD program package, the author make a research about the Tianwan nuclear power plant loading TVS-2M fuel assembly from the first cycle, also design the TVS-2M fuel assembly and on this basis, study fuel management, obtaining three fuel management cases, including year fuel cycle case and two long fuel cycle cases. In each program, the important parameters of the reactor core are analyzed and all the safety parameters meet the design requirements. In long fuel cycle program, TVS-2M is using from the first cycle and after the transition of only two cycles, the length of cycle reached the requirement of long period. The increased average annual capacity factor of the plant and the decreased times of overhaul during the core's life which saving 30.8 percentage of the overhaul cost due to the long fuel cycle can largely improve the economic efficiency of the plant. (authors)

  9. Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

  10. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  11. Performance Evaluation of Fuel Assembly Simulator for SMART

    International Nuclear Information System (INIS)

    Youn, Young Jung; Ko, Yung Joo; Chu, In Cheol; Euh, Dong Jin; Shin, Yong Cheol; Cho, Young Il; Kwon, Tae Soon

    2010-01-01

    SMART reactor has been being developed by KAERI for the generation of electric power and for seawater desalination. An experimental facility, called SCOP, is being constructed in order to evaluate the flow and pressure distribution in the SMART reactor core. The SCOP facility has a 1/5 linear scale of the prototype. The flow distribution at the inlet of 57 fuel assemblies will be measured at SCOP, and the experimental results will be used to evaluate the core thermal margin of SMART reactor. Each fuel assembly of SMART reactor will be simulated by single flow channel in the SCOP test facility. The fuel assembly simulator has a venturi tube at the front part to measure the flow rate through the channel, and several perforated plates to preserve the total pressure drop of the SMART fuel assembly. A CFD analysis was carried out to draw basic design parameters of the venturi tube and the perforated plates in the fuel assembly simulator. In the present work, the actual performance of the fuel assembly simulator was evaluated using a CALIP (Calibration Loop for Internal Pressure drop) test facility

  12. Bimetallic spacer means for a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1981-01-01

    A bimetallic spacer means designed to be cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The subject bimetallic spacer means in accord with one embodiment of the invention includes a member formed, at least principally, of Zircaloy to which are attached a plurality of stainless steel strips. The latter stainless steel strips are located on the external surface of the Zircaloy member and with the major axis of each of the plurality of stainless steel strips extending substantially perpendicular to the major axis of the Zircaloy member. In accord with another embodiment of the invention, the subject bimetallic spacer means includes a member formed at least principally of Zircaloy to which a plurality of stainless steel strips are attached so as to be positioned thereon externally thereof and with the major axis of each of the plurality of stainless steel strips extending substantially parallel to the major axis of the Zircaloy member. In accord with a further embodiment of the invention, the stainless steel strips are attached to preselected members, each embodying at least a cladding of Zircaloy, which are located in the rows of fuel rods that define the perimeter of the fuel matrix of the nuclear fuel assembly. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. Namely, the stainless steel strips expand laterally relative to the fuel assembly and thereby occupy the space adjacent to the external surface of the fuel assembly

  13. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    Yates, J.; Long, J.W.

    1975-01-01

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

  14. Results of trial operation of the WWER advanced fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

    2001-01-01

    The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

  15. FBFC's gadolinium fuel assembly manufacturing experience

    International Nuclear Information System (INIS)

    Van Den Eynde, M.; Belvegue, P.

    1999-01-01

    The burnable poison used by Framatome is gadolinium oxide integrated in the pellet by blending with UO 2 . This is the integrated poison which provides the largest experience feedback world-wide. Its main advantages are design flexibility and its well-known rod in reactor behaviour. FBFC's manufacturing experience with gadolinium is extensive. The first pellets were produced in 1986, present production averages 10 tons/year and cumulated experience reaches 47 tons. In parallel Framatome acquired gadolinium irradiation experience with more then 2 000 fuel assemblies in 33 reactors in 5 countries. Taking into account the increasing needs, a new gadolinium shop has been implemented in the FBFC Dessel plant. This shop, with a production capacity of 30 tU/yr is to be commissioned in the second quarter of 1999. It implements the most recent technological developments to achieve top product quality, safety and environment protection. (authors)

  16. Fuel assembly for pressure tube type reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Toshio.

    1990-01-01

    For providing a effect of burnable poisons without worsening local power peaking, it is effective to dispose conventional fuel rods incorporated with burnable poisons uniformly to the outermost layer of a fuel assembly in which fuel rods are disposed concentrically in a multilayered state. However, since the irradiation is applied to the outermost layer with high neutron fluxes in this case, the burnable poisons are eliminated rapidly and excess reactivity suppression effect does not last sufficiently. Then, according to the present invention, pellets of a dual layer structure comprising a pellet containing burnable poisons disposed at the center and nuclear fuel materials composed of plutonium-uranium mixed oxides or uranium oxides coated therearound are disposed. In view of the above, it is possible to obtain a fuel assembly which does not increase the suppression of excess reactivity at the initial stage and does not promote the elimination of burnable poisons, without lowering the local power peaking. (T.M.)

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

    International Nuclear Information System (INIS)

    Wu Zhongwang; Zhang Yajun

    2000-01-01

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

  18. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    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

  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. Equations of macrotransport in reactor fuel assemblies

    International Nuclear Information System (INIS)

    Sorokin, A.P.; Zhukov, A.V.; Kornienko, Yu.N.; Ushakov, P.A.

    1986-01-01

    The rigorous statement of equations of macrotransport is obtained. These equations are bases for channel-by-channel methods of thermohydraulic calculations of reactor fuel assemblies within the scope of the model of discontinuous multiphase coolant flow (including chemical reactions); they also describe a wide range of problems on thermo-physical reactor fuel assembly justification. It has been carried out by smoothing equations of mass, momentum and enthalpy transfer in cross section of each phase of the elementary fuel assembly subchannel. The equation for cross section flows is obtaind by smoothing the equation of momentum transfer on the interphase. Interaction of phases on the channel boundary is described using the Stanton number. The conclusion is performed using the generalized equation of substance transfer. The statement of channel-by-channel method without the scope of homogeneous flow model is given

  1. Methods of RECORD, an LWR fuel assembly burnup code

    International Nuclear Information System (INIS)

    Skardhamar, T.; Naess, H.K.

    1982-06-01

    The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)

  2. Fuel cycle cost evaluation to burnable absorber and feed assemblies

    International Nuclear Information System (INIS)

    Bae, S. M.; Lee, D. J.; Choi, H.

    1999-01-01

    The comparative analysis of fuel cycle cost for 16 equilibrium cores was performed to evaluate the fuel cycle cost depending on the combination of four different types of burnable absorbers (BAs) and feed assemblies for 18 month cycle operation of Westinghouse type 3-loop plant using 17*17 Vantage 5H fuel. The BAs considered are IFBA, Gadolinia, Erbia, and WABA and four cases of number of feed assemblies are 56, 60, 64, and 68. The optimal and practical combination of BA and number of feed assemblies is proposed. The optimum equilibrium core loading patterns for 16 cases were generated by using self-generating method and core design characteristics and the fuel cycle cost were evaluated for each loading pattern. The cycle lengths for all the cases were fixed as 480 EPFD to exclude effect of capital cost. Uranium ore cost, conversion cost, enrichment cost, fabrication cost, and back-end cycle cost were considered to evaluate fuel cycle cost. When considering the front-end cycle cost only, with the assumption of 8% discount rate, the 60 feed assemblies with using WABA case shows best result from the economic viewpoint. But the cost differences among the cases of using WABA, IFBA, and Gadolinia were negligible. And it was founded that the major factor for fuel cycle cost is BA price. In view of the core design characteristics, IFBA and Erbia were found to be more flexible than WABA and Gadolinia

  3. Fuel cycle cost evaluation to burnable absorber and feed assemblies

    International Nuclear Information System (INIS)

    Bae, Sung Man; Kim, Yun Ho; Shin, Ho Chul

    2001-01-01

    The comparative analysis of fuel cycle cost for 16 equilibrium cores was performed to evaluate the fuel cycle cost depending on the combination of four different types of burnable absorbers (BAs) and feed assemblies for 18 month cycle operation of Westinghouse type 3-loop plant using 17X17 Vantage 5H fuel. The BAs considered are IFBA, Gadolinia, Erbia, and W ABA and four cases of number of feed assemblies are 56, 00, 64, and 68. The optimal and practical combination of BA and number of feed assemblies is proposed. The optimum equilibrium core loading patterns for 16 cases were generated by using self-generating method and core design characteristics and the fuel cycle cost were evaluated for each loading pattern. The cycle lengths for all the cases were fixed as 480 EPFD to exclude effect of capital cost. Uranium ore cost, conversion cost, enrichment cost, fabrication cost, and back-end cycle cost were considered to evaluate fuel cycle cost. When considering the front-end cycle cost only, with the assumption of 8% discount rate, the 60 feed assemblies with using WABA case shows best result from the economic viewpoint. But the cost differences among the cases of using WABA, IFBA, and Gadolinia were negligible. And it was founded that the major factor for fuel cycle cost is BA price. In view of the core design characteristics, IFBA and Erbia were found to be more flexible than WABA and Gadolinia

  4. Top Nozzle Holddown Spring Optimization of KSNP Fuel Assembly

    International Nuclear Information System (INIS)

    Lee, Seong Ki; Park, Nam Kyu; Kim, Hyeong Koo; Lee, Joon Ro; Kim, Jae Won

    2002-01-01

    Nuclear fuel assembly for Korea Standard Nuclear Power (KSNP) Plant has 4 helical compression springs at the upper end of it. The springs, in conjunction with the fuel assembly weight, apply a holddown force against excess of buoyancy forces and the upward hydraulic forces due to the reactor coolant flow. Thus the holddown spring is to be designed such that the positive net downward force will be maintained for all normal and anticipated transient flow and temperature conditions in the nuclear reactor. With satisfying these in-reactor requirements of the fuel assembly holddown spring. Under the assumption that spring density is constant, the volume nozzle holddown spring. Under the assumption that spring density is constant, the volume minimization is executed by using the design variables, viz., wire diameter, mean coil diameter, minimization is executed by using the design variables, viz., wire diameter, mean coil diameter are within the compatible range of the fuel assembly structural components. Based on these conditions, the optimum design of the holddown spring is obtained considering the reactor operating condition and by using ANSYS code. The optimized spring has the properties that are a decreased volume and increased stiffness, compared with the existing one even if the absolute values are very similar each other. The holddown spring design features and the algorithm developed in this study could be directly applicable to the current commercial production. Therefore, it could be used to enhance the design efficiency and the functional performance of the spring, and to reduce a material cost a little

  5. 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.

  6. New phenomena observed during fuel assemblies testing

    International Nuclear Information System (INIS)

    Tzotcheva, V.

    2001-01-01

    The paper presents a new attempt to explain inexplicable increase of specific activity for some of the fuel assemblies during the fuel tightness testing procedures on Kozloduy NPP. A brief description of established procedure for fuel tightness control is presented in the paper. Special emphasis is given on a hypothesis that explains the fact of existence of deviation in Iodine activity more than usual, which have no reasonable interpretation. The reasons for uniform high Iodine activity for reloaded assemblies, that have kept in the open measuring can for a long time (1-3 hours), is found to be the process of Iodine dissolving in the water and the accelerated process of natural degassing. A proposal to use the 134 Cs and 137 Cs as stand-alone criteria for more precise results is made in respect to increase the reliability of fuel reloading and storage procedures

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-07-11

    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.

  9. Removing device for fuel assembly crud

    International Nuclear Information System (INIS)

    Ozaki, Shinji; Kanehira, Yoshinori; Iwai, Takashi.

    1993-01-01

    The device of the present invention removes cruds deposited on the surface of fuel assemblies in fuel pool water efficiently, to prevent leakage of cruds into the pool water. That is, a fuel cleaning vessel is disposed in the fuel pool water. Cruds on the surface of the fuel assembly are removed by a ultrasonic wave oscillator while the fuel assembly moves vertically in the cleaning vessel. Removed cruds are sucked together with water in the cleaning vessel by a submerged pump, and sent to a preceding coarse filter, in which relatively great cruds are captured. Fine cruds passing through the preceding filter are sent to a succeeding fine filter together with the filtered water by way of a water supply pump at the outside of the pool, in which fine cruds are completely captured. The filtered water is released into the pool water again. Since the two stage filters are prepared, each of the filters is less clogged. Accordingly, frequency for the exchange of the filters can be reduced. (I.S.)

  10. Corrosion of fuel assembly materials

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Beslu, P.

    1985-08-01

    Corrosion of zircaloy-4 is reviewed in relation with previsions of improvement in PWRs performance: higher fuel burnup; increase coolant temperature, implying nucleate boiling on the hot clad surfaces; increase duration of the cycle due to load-follow operation. Actual knowledge on corrosion rates, based partly on laboratory tests, is insufficient to insure that external clad corrosion will not constitute a limitation to these improvements. Therefore, additional testing within representative conditions is felt necessary [fr

  11. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

    Ashton, M.W.

    1975-01-01

    Reference is made to fuel element sub-assemblies for use in a Na cooled fast reactor. Such sub-assemblies may comprises a hexagonal bundle of slender fuel elements enclosed in a tubular sleeve, often referred to as a 'wrapper'. The fuel elements are spaced apart by helical wire wraps forming fins and which also space the wrapper from the bundle. The wire wraps make contact with the sheaths of adjacent elements and with the wrapper, so that each fuel element is well supported against thermal bowing, rattling and vibration, whilst allowing adequate coolant flow passages through the bundle. It has been found, however, that the outer fuel elements of the bundle are subject to over-cooling in this arrangement; this problem can, however, be largely overcome by reducing the flow passage between the bundle and the wrapper. In the arrangement described a wire filler is employed, extending along each outer coolant flow passage, and constructed in wave form. Fillers of such form have been found to reduce over-cooling considerably and they avoid the need for varied height wraps on the fuel elements. The fuel elements also have improved lateral support by contact with the fillers. (U.K.)

  12. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Ferrari, H.M.; Miller, D.L.; Tong, L.S.

    1975-01-01

    A description is given of a fuel assembly including multiple open channel grids for holding fuel rods and control rod guide thimbles in predetermined fixed relationship with each other. Metallic straps are interwoven to form a grid or egg crate configuration having openings which receive the fuel rods and guide thimbles. To properly support and cool the fuel rods near the grid-fuel rod interface, springs and dimples on the grid straps project into each opening, the dimples being oriented in a direction to permit flow of coolant upwardly therethrough. To minimize turbulence in coolant flow, the leading edge of each grid strap is provided with cutout sections which form scallops effective in channeling coolant in a uniform flow path through the network of grid openings

  13. CFD Analysis for a Fuel Assembly of GRR-1

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Park, Cheol; Kim, Heon Il; Chae, Hee Taek

    2010-06-15

    The thermal-hydraulic analysis was conducted on the research reactor core for improvement on the primary cooling system of GRR(Greece Research Reactor)-1. In order to design a primary cooling system, key data were provided by the thermal-hydraulic analysis. The COOLOD code was employed to carry out the thermal-hydraulic analysis, but it was for one-dimensional calculation and single channel analysis. It can't reproduce the three-dimensional flow in complex geometries. Although pressure drop through the fuel assembly was one of the most important values to design the primary cooling system, there was no data of it from an experiment or an estimation. It should be certain that the flow distribution between coolant channels was even, since all coolant channels of a plate type fuel assembly were completely separated from each other. However, those can be obtained by conducting an experiment, a quite long time and financial resources contribute to preventing an experiment. Regarding these, the CFD (Computational Fluid Dynamics) method was a very useful alternative to reach a solution to these problems. The CFD method provide reliable and useful predictions instead of experiments due to its applicability to complex shapes which were as real as possible. This is a summary report of CFD analysis for a plate type fuel assembly of GRR-1. In this study, flow distribution between each coolant channel of the fuel assembly was predicted. In order to estimate the pressure drop through the fuel assembly, many calculations were done for various flow rate conditions. A correlation between pressure drop to flow rate was yielded from those calculation results. Temperature distribution was estimated on the fuel plates of assembly at normal operation, and was compared with the prediction results obtained by the COOLOD code. Finally, it was predicted whether or not the uncovered core can be maintained under the core melting point only by air cooling of natural circulation, when the

  14. MOX fuel design and development consideration

    International Nuclear Information System (INIS)

    Yamate, K.; Abeta, S.; Suzuki, K.; Doi, S.

    1997-01-01

    Pu thermal utilization in Japan will be realized in several plants in late 1990's, and will be expanded gradually. For this target, adequacy of methods for MOX fuel design, nuclear design, and safety analysis has been evaluated by the committee of competent authorities organized by government in advance of the licensing application. There is no big difference of physical properties and irradiation behaviors between MOX fuel and UO 2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO 2 fuel will be applied with some modifications, taking into account of characteristic of MOX fuel. For nuclear design, new code system is to be applied to treat the heterogeneity in MOX fuel assembly and the neutron spectrum interaction with UO 2 fuel more accurately. For 1/3 MOX fueled core in three loop plant, it was confirmed that the fuel rod mechanical design could meet the design criteria, with slight reduction of initial back-fitting pressure, and with appropriate fuel loading patterns in the core to match power with UO 2 fuel. With the increase of MOX fuel fraction in the core, control rod worth and boron worth decrease. Compensating the decrease by adding control rod and utilizing enriched B-10 in safety injection system, 100% MOX fueled core could be possible. Up to 1/3 MOX fueled core in three loop plant, no such modifications of the plant is necessary. The fraction of MOX fuel in PWR is designed to less than 1/3 in the present program. In order to improve Pu thermal utilization in future, various R and D program on fuel design and nuclear design are being performed, such as the irradiation program of MOX fuel manufactured through new process to the extent of high burnup. (author). 8 refs, 9 figs, 2 tabs

  15. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.; Ashton, M.W.; Aubertin, J.C.

    1975-01-01

    Reference is made to fuel element sub-assemblies for use in a Na cooled fast reactor. Such sub-assemblies may comprise a hexagonal bundle of slender fuel elements enclosed in a tubular sleeve, often referred to as a 'wrapper'. The fuel elements are spaced apart by helical wire wraps forming fins and which also space the wrapper from the bundle. The wire wraps make contact with the sheaths of adjacent elements and with the wrapper, so that each fuel element is well supported against thermal, bowing, rattling and vibration, whilst allowing adequate coolant flow passages through the bundle. The arrangement of the fuel elements usually provides two groups of passageways for coolant flow through the bundle -an inner group each bounded by four fuel elements, and an outer group each bounded by two outer fuel elements and the wrapper. It has been found, however, that this arrangement results in over-cooling of the outer fuel elements. An arrangement is described that overcomes this disadvantage to a considerable extent. The fuel elements are arranged in three groups - an inner group, and intermediate group and an outer group. The fins on the outer group elements are helically wound at half the pitch of the other two groups and alternate outer group elements are wound opposite handed to abutting outer row elements; elements of the outer and intermediate groups are orientated and end located to enable them to nest together. By utilising half pitch helical fins on the outer elements a greater number of lateral support points is obtained for the outer and intermediate groups of elements and the increased number of turns of fins on the outer elements assist in reducing overcooling of the outer elements. (U.K.)

  16. nuclear fuel design criteria

    International Nuclear Information System (INIS)

    Can, S.

    1997-01-01

    Nuclear fuel design is strictly dependent on reactor type and experiences obtained from performance of nuclear fuels. The objectives of the design are reliability, and economy. Nuclear fuel design requires an interdisciplinary work which has to cover, at least nuclear design, thermalhydraulic design, mechanical design, and material properties.The procedure of design, as describe in the quality assurance, consist of a number of steps. The most important parts are: Design description or inputs, preliminary design, detailed design and design output, and design verification. The first step covers objectives and requirements, as defined by the customer and by the regulatory authority for product performance,environmental factors, safety, etc. The second describes assumptions and alternatives, safety, economy and engineering analyses. The third covers technical specifications, design drawings, selection of QA program category, etc. The most important form of design verification is design review by qualified independent internal or external reviewers. The scope of the review depends on the specific character of the design work. Personnel involved in verification and review do not assume prime responsibility for detecting errors. Responsibility for the design remains with the personnel involved in the design work

  17. 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.

  18. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  19. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Billerey, Antoine; Waeckel, Nicolas

    2005-01-01

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

  1. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S.

    2012-07-24

    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.

  2. Inverse design of multicomponent assemblies

    Science.gov (United States)

    Piñeros, William D.; Lindquist, Beth A.; Jadrich, Ryan B.; Truskett, Thomas M.

    2018-03-01

    Inverse design can be a useful strategy for discovering interactions that drive particles to spontaneously self-assemble into a desired structure. Here, we extend an inverse design methodology—relative entropy optimization—to determine isotropic interactions that promote assembly of targeted multicomponent phases, and we apply this extension to design interactions for a variety of binary crystals ranging from compact triangular and square architectures to highly open structures with dodecagonal and octadecagonal motifs. We compare the resulting optimized (self- and cross) interactions for the binary assemblies to those obtained from optimization of analogous single-component systems. This comparison reveals that self-interactions act as a "primer" to position particles at approximately correct coordination shell distances, while cross interactions act as the "binder" that refines and locks the system into the desired configuration. For simpler binary targets, it is possible to successfully design self-assembling systems while restricting one of these interaction types to be a hard-core-like potential. However, optimization of both self- and cross interaction types appears necessary to design for assembly of more complex or open structures.

  3. Calibration Tests of Fuel Assembly Simulators of APR+ Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kih Wan; Chu, In Cheol; Euh, Dong Jin; Kwon, Tae Soon [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    A Reactor flow distribution is regarded to be major importance in improving the design margin of a flow distribution. The prediction of APR+ core fluid flow phenomena has been in demand, since 257 fuel assemblies are adapted in the APR+, unlike in the APR1400. The APR+ reactor flow test facility, the ACOP (APR+ Core Flow and Pressure Test Facility), was constructed to analyze the hydraulic characteristics. For the ACOP facility, the core simulator was designed with a scale analysis to simulate the real HIPER fuel assembly of an APR+. In this study, for all 257 core simulators, several calibration tests were conducted to verify their design performance before applying them to the ACOP facility. The inlet flow rate and the total pressure drop of the simulators were measured by varying flow rates to evaluate its compatibility. The discharge coefficients were also calculated from the experimental data to produce a statistical database for a further ACOP facility test.

  4. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.

    1977-01-01

    A method of securing a fuel bundle to permit easy remote disassembly is described. Fuel rods are held loosely between end plates, each end of the rods fitting into holes in the end plates. At the upper end of each fuel rod there is a spring pressing against the end plate. Tie rods are used to hold the end plates together securely. The lower end of each tie rod is screwed into the lower end plate; the upper end of each tie rod is attached to the upper end plate by means of a locking assembly described in the patent. In order to remove the upper tie plate during the disassembly process, it is necessary only to depress the tie plate against the pressure of the springs surrounding the fuel rods and then to rotate each locking sleeve on the tie rods from its locked to its unlocked position. It is then possible to remove the tie plate without disassembling the locking assembly. (LL)

  5. Removal and replacement of fuel rods in nuclear fuel assembly

    International Nuclear Information System (INIS)

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

    1983-01-01

    Apparatus for replacing components of a nuclear fuel assembly stored in a pit under about 10 m. of water. The fuel assembly is secured in a container which is rotatable from the upright position to an inverted position in which the bottom nozzle is upward. The bottom nozzle plate is disconnected from the control-rod thimbles by means of a cutter for severing the welds. To guide and provide lateral support for the cutter a fixture including bushings is provided, each encircling a screw fastener and sealing the region around a screw fastener to trap the chips from the severed weld. Chips adhering to the cutter are removed by a suction tube of an eductor. (author)

  6. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  7. 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.

  8. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  9. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  10. Fuel assembly for PWR type reactor

    International Nuclear Information System (INIS)

    Yokoyama, Takashi.

    1991-01-01

    In a fuel assembly in a reactor-loaded state, pellets to be loaded in a region higher than a predetermined height are made hollow. That is, the volume of the gap in the hollow portion of the pellet comprising fissible materials to be filled at a position higher by 1/2 to 1/3 height from the upper region (downstream of coolant flow) of at least a portion of fuel rods in a fuel assembly is reduced stepwise than that in the lower region (upstream of coolant flow). Alternatively, the volume of the gap is gradually reduced from the lower portion to the upper portion. The diameter of the hollow hole portion is thus varied and the volume of the gap of the fissible materials per pellet is controlled, to reduce the amount of the fissile materials per unit volume. With such a constitution, the power of fission energy in the reactor core upper portion can be lowered. Accordingly, the corrosion of a cladding tube upon high burnup degree can be suppressed, thereby enabling to ensure integrity. (T.M.)

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

    Science.gov (United States)

    Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

    2015-12-15

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

  12. Bottom nozzle of a LWR fuel assembly

    International Nuclear Information System (INIS)

    Leroux, J.C.

    1991-01-01

    The bottom nozzle consists of a transverse element in form of box having a bending resistant grid structure which has an outer peripheral frame of cross-section corresponding to that of the fuel assembly and which has walls defining large cells. The transverse element has a retainer plate with a regular array of openings. The retainer plate is fixed above and parallel to the grid structure with a spacing in order to form, between the grid structure and the retainer plate a free space for tranquil flow of cooling water and for debris collection [fr

  13. 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)

  14. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Sergi

    2012-01-01

    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

  15. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

  16. Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. Measurements of the natural frequencies for the HANARO fuel assembly

    International Nuclear Information System (INIS)

    Park, Jin Ho; Kim, Tae Ryong; Ryu, Jeong Soo

    1998-01-01

    Measurement of the natural frequencies for the fuel assembly mock-up of HANARO reactor, which is a research reactor operating in KAERI, wa implemented. There are two types of the fuel assembly model, one is 18-element fuel bundle assembly and the other is 36-element one. They were locked inside the shell type flow tubes, respectively. The flow tube (round flow tube) corresponding to the 18-element fuel assembly mock-up has a form of cylindrical shell and the tube (hexagonal flow tube) to the 36-element fuel assembly model does hexagonal shell. The in-air fundamental natural frequency of the round flow tube was turned out to be 54 Hz and the in-water one 26 Hz. The in-air fundamental natural frequency of the hexagonal flow tube resulted as 58.5 Hz and the in-water one was reduced to 29 Hz due to added mass effect. Also the in-air fundamental natural frequency of the 18-element fuel assembly structure (fuel assembly and round flow tube) was found to be 26 Hz and the in-water one 16 Hz. Finally the in-air frequency of the 36-element fuel assembly one (fuel assembly and hexagonal flow tube) was estimated as 28 Hz, and in-water one 11 Hz. (author). 3 refs., 2 tabs., 45 figs

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

    International Nuclear Information System (INIS)

    Billerey, A.; Bouffioux, P.

    2002-01-01

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

  20. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal

    International Nuclear Information System (INIS)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1

  1. Calibration of spent fuel measurement assembly

    Science.gov (United States)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-11-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system.

  2. An improved assembly for the transport of fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1979-01-01

    An improved assembly for the transport and storage of radioactive nuclear fuel elements is described. The fuel element transport canister is of the type in which the fuel elements are submerged in liquid with a self regulating ullage system, so that the fuel elements are always submerged in the liquid even when the assembly is used in one orientation during loading and another orientation during transportation. (UK)

  3. 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).

  4. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    1990-10-01

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  5. Neutronics assessment of thorium-based fuel assembly in SCWR

    International Nuclear Information System (INIS)

    Liu, Shichang; Cai, Jiejin

    2013-01-01

    Highlights: • A novel thorium-based fuel assembly for SCWR has been introduced and investigated. • Neutronic properties of three thorium fuels have been studied, compared with UO 2 fuel. • The thorium-based fuel has advantages on fuel utilization and lower MAs generation. -- Abstract: Aiming to take advantage of neutron spectrum of SCWR, a novel thorium-based fuel assembly for SCWR is introduced in this paper. The neutronic characteristics of the introduced fuel assembly with three different thorium fuel types have been investigated using the “dragon” codes. The parameters in different working conditions, such as infinite multiplication factors, radial power peaking factor, temperature coefficient of reactivity and their relation with the operation period have been assessed by comparing with conventional uranium assembly. Moreover, the moderator-to-fuel ratio (MFR) was changed in order to investigate its influence on the neutronic characteristics of fuel assembly. Results show that the thorium-based fuel has advantages on both efficient fuel utilization and lower minor actinide generation, with some similar neutronic properties to the uranium fuel

  6. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

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

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  7. Tomographic techniques for safeguards measurements of nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist Saleh, Tobias

    2007-10-15

    Nuclear power is currently experiencing increased interest over the world. New nuclear reactors are being built and techniques for taking care of the nuclear waste are being developed. This development puts new demands and standards to safeguards, i.e. the international efforts for ensuring the non-proliferation of nuclear weapons. New measuring techniques and devices are continuously being developed for enhancing the ability to detect diversion of fissile material. In this thesis, tomographic techniques for application in safeguards are presented. Tomographic techniques can non-destructively provide information of the inner parts of an object and may thus be used to control that no material is missing from a nuclear fuel assembly. When using the tomographic technique described in this thesis, the radiation field around a fuel assembly is first recorded. In a second step, the internal source distribution is mathematically reconstructed based on the recorded data. In this work, a procedure for tomographic safeguards measurements is suggested and the design of a tomographic measuring device is presented. Two reconstruction algorithms have been specially developed and evaluated for the application on nuclear fuel; one algorithm for image reconstruction and one for reconstructing conclusive data on the individual fuel rod level. The combined use of the two algorithms is suggested. The applicability for detecting individual removed or replaced rods has been demonstrated, based on experimental data

  8. 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.)

  9. Transient assembly of active materials fueled by a chemical reaction

    Science.gov (United States)

    Boekhoven, Job; Hendriksen, Wouter E.; Koper, Ger J. M.; Eelkema, Rienk; van Esch, Jan H.

    2015-09-01

    Fuel-driven self-assembly of actin filaments and microtubules is a key component of cellular organization. Continuous energy supply maintains these transient biomolecular assemblies far from thermodynamic equilibrium, unlike typical synthetic systems that spontaneously assemble at thermodynamic equilibrium. Here, we report the transient self-assembly of synthetic molecules into active materials, driven by the consumption of a chemical fuel. In these materials, reaction rates and fuel levels, instead of equilibrium composition, determine properties such as lifetime, stiffness, and self-regeneration capability. Fibers exhibit strongly nonlinear behavior including stochastic collapse and simultaneous growth and shrinkage, reminiscent of microtubule dynamics.

  10. Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

    International Nuclear Information System (INIS)

    Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

    1986-01-01

    Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

  11. 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.

  12. Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

    International Nuclear Information System (INIS)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

    2002-12-01

    In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

  13. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    International Nuclear Information System (INIS)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

    1999-01-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

  14. 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.

  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. 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

  17. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  18. Fuel radial design using Path Relinking

    International Nuclear Information System (INIS)

    Campos S, Y.

    2007-01-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  19. CORD, PWR Core Design and Fuel Management

    International Nuclear Information System (INIS)

    Trkov, Andrej

    1996-01-01

    1 - Description of program or function: CORD-2 is intended for core design applications of pressurised water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refuelling). 2 - Method of solution: The calculations are performed at the cell level with a lattice code in the supercell approximation to generate the single cell cross sections. Fuel assembly cross section homogenization is done in the diffusion approximation. Global core calculations can be done in the full three-dimensional cartesian geometry. Thermohydraulic feedbacks can be accounted for. The Effective Diffusion Homogenization method is used for generating the homogenized cross sections. 3 - Restrictions on the complexity of the problem: The complexity of the problem is selected by the user, depending on the capacity of his computer

  20. Fuel assembly gripping device using self-locking mechanism

    International Nuclear Information System (INIS)

    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

  1. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

    2011-01-01

    In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

  2. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  3. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    International Nuclear Information System (INIS)

    Bobrov, E.; Alekseev, P.; Chibinyaev, A.; Teplov, P.; Dudnikov, A.

    2016-01-01

    REMIX (Regenerated Mixture) fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessed used fuel and the fabrication technology of such fuel is called REMIX-technology. This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water-fuel ratio in the VVER FA affects the fuel characteristics produced by REMIX technology during multiple recycling. It is shown that for for the traditional REMIX-fuel it does not make sense to change anything in the design of VVER FA, because there are no advantages in the fuel feed consumption. The natural uranium economy by the fifth cycle reached about 29%. In the case of the REMIX fuel based on uranium-plutonium from SNF MOX fuel, it would be appropriate to use fuel assemblies with a water-fuel ratio of 1.5

  4. Inspection and repair apparatus for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Hornak, L.P.; Desmarchais, W.E.

    1975-01-01

    An apparatus is disclosed for inspecting and repairing a radioactive fuel assembly. The radioactive fuel assembly is positioned within a shielding sleeve which substantially reduces the level of radioactivity immediately surrounding the sleeve thereby permitting direct access by operating personnel. In one embodiment, a rotatable collar is mounted to the sleeve at a midlength location. An access port, an inspection port and an instrument port are included with the collar so that operating personnel may directly inspect the fuel assembly and effectuate any necessary repairs

  5. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-01-27

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC.

  6. 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.

  7. Support for a storage rack for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Friedrichs, H.; Heinz, G.; Krainer, F.; Swelim, H.; Eisner, J.

    1983-01-01

    For support of a storage rack with a rectangular cross-section for nuclear reactor fuel assemblies elbows to be used as centering elements may be fastened to the metal licencing of a storage pit by means of rectangular side pieces, which are arranged at the corners of the fuel pit in such manner that rectangular feet positioned there will be enclosed at the outside of the fuel pit. For adjustment of a given clearance the elbows may be provided with fitting pieces. They also may have supporting plates serving as bases for the feet. The invention may be of advantage especially for storing fuel assemblies from light water reactors. (orig./PW)

  8. Experience feedback from the transportation of Framatome fuel assemblies

    International Nuclear Information System (INIS)

    Robin, M.E.; Gaillard, G.; Aubin, C.

    1998-01-01

    Framatome, the foremost world nuclear fuel manufacturer, has for 25 years been delivering fuel elements from its three factories (Dessel, Romans, Pierrelatte) to the various sites in France and abroad (Germany, Sweden, Belgium, China, Korea, South Africa, Switzerland). During this period, Framatome has built up experience and expertise in fuel element transportation by road, rail and sea. In this filed, the range of constraints is very wide: safety and environmental protection constraints; constraints arising from the control and protection of nuclear materials, contractual and financial constraints, media watchdogs. Through the experience feedback from the transportation of FRAMATOME assemblies, this paper addresses all the phases in the transportation of fresh fuel assemblies. (authors)

  9. Grid spacers for use in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kuwako, Akira.

    1987-01-01

    Purpose: To obtain spacers capable of reducing the pressure loss by enlarging coolant flow channels when the fuel temperature is high, while capable of reliably maintaining the fuel pins with no vibrations when the fuel temperature is low. Constitution: This invention concerns grid spacers for constituting fuel assemblies for use in water cooled reactors. Memory shape alloys are disposed at least a portion of a spacer element that takes such a shape as urging the pin when the fuel temperature is low, while enlarging the coolant flow channel to reduce the pressure loss when the fuel temperature is high. (Ikeda, J.)

  10. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

  11. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

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

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

  12. Computational analysis of the effect of fuel assembly features on the core thermohydraulics at elevated parameters

    International Nuclear Information System (INIS)

    Shcherbakov, S.; Sergeev, V.

    2009-01-01

    The results of the numerical calculations for single- and two-phase coolant flows in the WWER-1000 core at elevated parameters are presented. The non-uniformities of flow in the core caused by design features of fuel assemblies (TVS-2 or TVSA, different number of spacers, use of mixing grids), boiling in some fuel assemblies with increasing power under non-uniform power generation and non-uniform hydraulic resistance are discussed. The stability limits of flow in the core are estimated. The non-uniformity of power generation in fuel assemblies or their hydraulic resistance was demonstrated to result in disproportionate intensive boiling in some fuel assemblies and in reduction of the margin to CHF. The calculations were performed using the TURBOFLOW 2D code. It was shown that there are no important differences in the use of only TVS-2 or TVSA in the core (authors)

  13. Simulating control rod and fuel assembly motion using moving meshes

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, D. [Department of Electrical and Computer Engineering, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)], E-mail: gilbertdw1@gmail.com; Roman, J.E. [Departamento de Sistemas Informaticos y Computacion, Universidad Politecnica de Valencia, Camino de Vera s/n, 46022 Valencia (Spain); Garland, Wm. J. [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada); Poehlman, W.F.S. [Department of Computing and Software, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)

    2008-02-15

    A prerequisite for designing a transient simulation experiment which includes the motion of control and fuel assemblies is the careful verification of a steady state model which computes k{sub eff} versus assembly insertion distance. Previous studies in nuclear engineering have usually approached the problem of the motion of control rods with the use of nonlinear nodal models. Nodal methods employ special approximations for the leading and trailing cells of the moving assemblies to avoid the rod cusping problem which results from the naive volume weighted cell cross-section approximation. A prototype framework called the MOOSE has been developed for modeling moving components in the presence of diffusion phenomena. A linear finite difference model is constructed, solutions for which are computed by SLEPc, a high performance parallel eigenvalue solver. Design techniques for the implementation of a patched non-conformal mesh which links groups of sub-meshes that can move relative to one another are presented. The generation of matrices which represent moving meshes which conserve neutron current at their boundaries, and the performance of the framework when applied to model reactivity insertion experiments is also discussed.

  14. Nuclear reactor fuel assemblies and end fitting grid structures therefor

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    An improved end fitting grid structure is described for nuclear fuel assemblies which overcomes the need for load-bearing control rod guide tubes and the expensive special fittings that these tubes required. (UK)

  15. Support a nuclear fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Leclercq, J.

    1985-01-01

    The device has to maintain the assemblies with regard to a horizontal plate of the core. The assemblies, having the same section, are arranged side by side in a regular polygonal lattice and each asssembly is, either equipped with at least two zones to receive the rods which are vertically inserted and maintained during the reactor operation, or beside an assembly which is equipped. The device has two sets comprising each one at least one deformable locking element and a rigid element which raches with it, one fixed to the fuel assembly and the other fixed to a horizontal plate attached to the reactor core, positioned so that inserting a fuel rod into an emplacement in the fuel assembly deforms the bolt transversally to lock it with the rigid piece. The invention can be applied to water moderated reactors [fr

  16. 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.

  17. Monitoring and Leak testig of wwer-440 fuel assemblies in Slovak wet interim spent fuel storage facility

    Directory of Open Access Journals (Sweden)

    Miroslav Božik

    2007-01-01

    Full Text Available An accelerated monitoring system designed for the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice bases on the newly designed “cesium detectors” is presented in the paper. Since 1999, leak tests of WWER-440 fuel assemblies are provided by a special leak tightness detection system “Sipping in Pool” delivered by the Framatome-anp with external heating for the precise defects determination. Although this system seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long. Therefore, a new “on-line” detection system, based on the new sorbent NIFSIL for an effective 134Cs and 137Cs activity was developed. The design of this detection system and its application possibility in Slovak wet interim spent fuel storage facility as well as preliminary results are presented.

  18. Design of the MYRRHA Spallation Target Assembly

    International Nuclear Information System (INIS)

    Keijers, S.; Fernandez, R.; Stankovskiy, A.; Kennedy, G.; Van Tichelen, K.

    2015-01-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is a multi-purpose research facility currently being developed at SCK.CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level. As a flexible irradiation facility, the MYRRHA research reactor will be able to work in both critical and subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material research for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by Lead Bismuth Eutectic (LBE) and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. This paper describes the evolution of the MYRRHA spallation target design. In the early phase of the MYRRHA project (XT-ADS), the target design was based on a dedicated spallation loop inside the primary reactor vessel. Within the core, the 3 central fuel assembly positions were occupied by the spallation target, which enabled a windowless design created by a free surface of LBE facing the proton beam. The windowless option was preferred because of high heat loads in combination with severe irradiation damage in the target region would result in unacceptably short lifetimes of a target window. The LBE in the loop served as spallation target and as target coolant, but was separated from the LBE cooling the reactor core. The loop was equipped with its own pump, heat exchanger and conditioning system. The change from cyclotron to linear accelerator allowed the increase in proton energy from 350 MeV to 600 MeV. This modification led to an important reduction of the specific heat load at the target level and an improvement of the neutronic performance. In addition to

  19. Fabrication details for wire wrapped fuel assembly components

    International Nuclear Information System (INIS)

    Bosy, B.J.

    1978-09-01

    Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

  20. Advanced fuel assemblies for economic and flexible operation of light water reactors

    International Nuclear Information System (INIS)

    Urban, P.; Bender, D.

    2001-01-01

    Increasing competition in the electricity market sets up a corresponding competition between the different electricity producing technologies. This makes further improvements in the economics of nuclear power generation a vital item for the future of nuclear energy. Though the costs for development, design and fabrication of fuel assemblies contribute only about 10% to the fuel cycle costs, the design and the performance of the fuel assemblies considerably influences total electricity generation cost. By the recent creation of Framatome ANP the nuclear activities of Framatome and Siemens were combined into one company. In the past, both had made considerable achievements in the development of fuel assemblies and related services supporting the goal of safe and economic electricity generation by light water reactors. The examples described in this paper cover former Siemens products and experience. In the future, our combined experience bases will be an ideal platform to offer further substantial improvements to our customers. (author)

  1. 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.

  2. Equipment designs for the spent LWR fuel dry storage demonstration

    International Nuclear Information System (INIS)

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

  3. Tomographic imaging of severely disrupted fuel assemblies tested in TREAT

    International Nuclear Information System (INIS)

    Morman, J.A.; Froehle, P.H.; Holland, J.W.; Bennett, J.D.

    1990-01-01

    A series of CT codes is under development in the Reactor Analysis and Safety Division of Argonne National Laboratory for use as a post-test examination tool to analyze segments of the final fuel-bundle configuration of TREAT tests. This paper presents the results of CT analysis for fuel assemblies using neutron radiography. Fuel relocation following overpower transients in the TREAT reactor is examined for sections of the assemblies, and results are compared to metallographic sections. Further improvements are expected to increase the use and reliability of CT analysis as a standard post-test examination tool

  4. An improved assembly for the transport of fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1979-01-01

    An improved assembly is described for the transport and storage of radioactive fuel elements. The system consists of a transport flask in which the fuel element holder is placed in such a way that the elements may be submerged in liquid within the flask and the assembly used in one orientation for loading the fuel elements and in another orientation for transporting them. The assembly has a self-regulating ullage system comprising reservoirs for containing liquid and pressurising gas, the arrangement for the reservoirs being such that in either orientation of the assembly liquid is maintained in all the reservoirs to prevent egress of the pressurised gas and to compensate for volume changes arising from temperature variations within the flask. (U.K.)

  5. Stress Linearization and Strength Evaluation of the BEP's Flow Plates for a Dual Cooled Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Kang, Heung Seok; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2009-01-01

    A fuel assembly is composed of 5 major components, such as a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT) and fuel rods (FRs). There are no ASME criteria about all components except for a TEP/BEP. The TEP/BEP should satisfy stress intensity limits in case of condition A and B of ASME, Section III, Division 1 . Subsection NB. In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape to provide flow holes to direct the heated coolant into/out of the fuel assembly but structural intensity of these plates within a 22.241 kN axial loading should satisfy Tresca stress limits in ASME code. In this paper, stress linearization procedure and strength evaluation of a newly designed BEP for the dual cooled fuel assembly are described

  6. Nuclear Fuel Assembly Assessment Project and Image Categorization

    International Nuclear Information System (INIS)

    Lindsey, C.S.; Lindblad, T.; Waldemark, K.; Hildingsson, Lars

    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

  7. 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.

  8. Design package for vacuum wand for fuel retrieval system

    International Nuclear Information System (INIS)

    ROACH, H.L.

    1999-01-01

    This is a design package that contains the details for the design, fabrication, and testing of a vacuum wand that will pick up sludge and corrosion products generated during fuel assembly handling operations at K-Basin. This document contains requirements, development design information, design calculations, tests, and test reports

  9. Contemporary and prospective fuel cycles for WWER-440 based on new assemblies with higher uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    Gagarinskiy, A.A.; Saprykin, V.V.

    2009-01-01

    RRC 'Kurchatov Institute' has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for WWER-440 reactors. Works were performed to upgrade and improve WWER-440 fuel cycles on the basis of second-generation fuel assemblies allowing core thermal power to be uprated to 107 108 % of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87 % of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of WWER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of WWER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (Authors)

  10. 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.

  11. Review of qualifications for fuel assembly fabrication

    International Nuclear Information System (INIS)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian

    2013-01-01

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  12. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  13. Application of ultra-sons to on-site spent fuel assemblies metrology

    International Nuclear Information System (INIS)

    Gondard, C.; Saglio, R.; Vouillot, M.; Delaroche, P.; Vaubert, Y.; Van Craeynest, J.C.

    1983-12-01

    Fuel assemblies inspection on the site of a power reactor, between two irradiation campaigns, allows to estimate the behaviour of prototype fuel assemblies and to permit their refueling for the continuation of the irradiation; the utilization of non-destructive, reliable and high-performance techniques, is of a great interest in the application. For, this reason, the C.E.A. has been led to carry out new techniques allowing the visual examination and the dimensional inspection of spent fuel assemblies of 900 MWe French pressurized water reactors, with a transportable Fuel Examination Module (MEC) on every reactor site. This module includes a television camera, and uses for the first time as ''position sensor'' the properties offered by a set of ultrasonic transducers. The main principle of the design, of the operation way of the module, of the measuring methods, and, of the data acquisition and processing, are presented [fr

  14. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2005-01-01

    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)

  15. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  16. 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.

  17. Development of packagings for 'MONJU' blanket fuel assemblies

    International Nuclear Information System (INIS)

    Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

    1995-01-01

    Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

  18. Trends in joining technology development for PHWR fuel assemblies in India

    International Nuclear Information System (INIS)

    Desai, P.B.; Kulkarni, P.G.

    1999-01-01

    Various types of welding/brazing techniques and equipment are used in the production of different kinds of PHWR fuel assemblies. These assemblies are of 19-element, 22-element, 28-element and 37-element type and are unique in joint design. All joints in PHWR fuel assemblies are either welded or brazed without any mechanical joints. Resistance welding, GTAW, Electron Beam welding, brazing and ultrasonic welding are some of the feasible techniques. Resistance welding and brazing are extensively used in fabrication of fuel assemblies involving joining of Zirconium alloys components of different shape and geometry. This paper describes recent advances made and work caused out at Atomic Fuels Division of Bhabha Atomic Research Centre in India in fabrication of different types of PHWR fuel assemblies. Vast experience gained in development of technologies and large-scale production of different core components was utilised to evaluate, develop and assess joining techniques for production. Various joining processes were utilised to achieve improved reliability and economy of products. Several non-conventional processes of joining have been under consideration. Modern Developments in welding power sources, methods of achieving improved weld quality and consistency, advances in weld monitoring etc are described. Efforts made towards better joint quality, ease of production and improved fuel performance is discussed Paper includes work carried out successfully in indigenous fabrication of some equipment, automation of processes and monitoring system. (author)

  19. One approach to accepting and transporting spent fuel from early-generation reactors with short fuel assemblies

    International Nuclear Information System (INIS)

    Peterson, R.W.; Bentz, E.J. Jr.; Bentz, C.B.

    1993-01-01

    In the early days of development of commercial nuclear power reactors in the U.S., the overall length and uranium loading of the fuel assemblies were considerably less than those of later generation facilities. In turn, some of these early facilities were designed for handling shorter casks than currently-certified casks. The spent fuel assemblies from these facilities are nearly all standard fuel within the definition in the Standard Contract (10 CFR 961) between the utilities and the U.S. Department of Energy (DOE) (the Big Rock Point fuel cross-section is outside the standard fuel dimension), and the utilities involved hold early delivery rights under DOE's oldest-fuel-first (OFF) allocation scenario. However, development of casks suitable for satisfying the acceptance and transportation requirements of some of these facilities is not currently underway in the DOE Cask System Development Program (CSDP). While the total MTU of these fuels is relatively small compared to the total program, the number of assemblies to be transported is significant, especially in the early years of operation according to the OFF allocation scenario. We therefore perceive a need for DOE to develop an approach and to implement plans to satisfy the unique acceptance and transportation requirements of these facilities. One such approach is outlined below. (author)

  20. Stress analysis of fuel assemblies under seismic load

    International Nuclear Information System (INIS)

    Kiselev, A.; Krutko, E.; Kiselev, I.; Tutnov, A.

    2011-01-01

    One of the important parts of fuel assemblies (FA) safety validation is their strength estimation under the dynamic loads, such as the vibration effects caused by the work of reactor units and the seismic exposure of an earthquake, leading to extreme inertia loads on all elements of the NPP. Taking into account structural features of FA and a very large mass, the exposure of seismic loads can lead to significant deformation of fuel assemblies. It is necessary to assess the magnitude of the force interaction between the FA in case of an earthquake to estimate the strength and performance of fuel assemblies. It is also necessary to compute FA bending forms and maximum values for further RPS control rods inserting time estimation, and for disassembly possibility justification of the core and individual FA after the earthquake. The problem of WWER-1000 core dynamic behavior modeling with TVS-2M fuel assemblies under the seismic loads exposure using the finite element method is described. Each fuel assembly is represented by equivalent rod finite element model. The reactor core is simulated by 163 fuel assemblies in accordance with the reactor core construction. Stiffness characteristics of fuel assemblies are determined on the results of a series of static and dynamic TVS-2M FA field tests. The special algorithm was developed to consider the fuel rod slippage effect during deformation. The special contact elements are introduced into the model of the core to take into account the interaction of fuel assemblies with their neighbors and with core barrel. Solution of the dynamic equilibrium equations system of finite element model is implemented by direct integration using the explicit scheme. Parallel algorithms for numerical integration on multiprocessor computers with graphics processing unit is developed to improve the efficiency of calculations. Values of nodes displacement in finite element model of reactor core as a function of seismic excitation time are obtained

  1. Preliminary design of smart fuel

    International Nuclear Information System (INIS)

    Kim, Y.; Ha, D.; Park, S.; Nahm, K.; Lee, K.; Kim, J.

    2007-01-01

    SMART (System-integrated Modular Advanced Reactor) is a novel light water rector with a modular, integral primary system configuration. This concept has been developing a 660 MWt by Korean Nuclear Power Industry Group with KAERI. SMART is being developed for use as an energy source for small-scale power generation and seawater desalination. Although the design of SMART is based on the current pressurized water reactor technology, new technologies such as enhanced safety, and passive safety have been applied, and system simplification and modularization, innovations in manufacturing and installation technologies have been implemented culminating in a design that has enhanced safety and economy, and is environment -friendly. In this paper described the preliminary design of the nuclear Fuel for this SMART, the design concept and the characteristics of SMART Fuel. In specially this paper describe the optimization of grid span adjustment to improve the thermal performance of the SMART Fuel as well as to improve the seismic resistance performance of the SMART Fuel, it is not easy to improve the both performance simultaneously because of design parameter of each performance inversely proportional. SMART Fuel enable to extra-long extended fuel cycle length and resistance of proliferation, enhanced safety, improved economics and reduced nuclear waste

  2. 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.

  3. Fuel assemblies, grapples therefor and fuel transport apparatus for nuclear reactor power plant

    International Nuclear Information System (INIS)

    Jones, C.R.

    1975-01-01

    A description is given of a nuclear fuel assembly in which vertically disposed fuel elements are spaced within a housing generally of a rectanguloid configuration. Each fuel element includes an upper end plug and lower end plug. Vertically spaced support plates are disposed in the housing with suitable openings to receive the upper and lower end plugs of the fuel elements for supporting the fuel elements with the housing. The upper plate is removable from the housing and the lower fuel plug is detachably connected to the lower plate. Other spacer plates are secured to the housing walls to reinforce same. A grapple having lifting plates with pins enters recesses formed in the housing for enabling the housing to be raised. After the fuel assembly is raised by the grapple, leaf spring retainers of the upper plate are dislodged for removing the upper plate from the housing. Now, the fuel elements can be removed selectively and individually from the fuel assembly by a removal tool. Aligned with and disposed above the removal tool is a transfer casing for housing the selectively removed fuel element while the selectively removed fuel element is transported to and from a fuel reprocessor

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

    International Nuclear Information System (INIS)

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

    1984-10-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. Device for supporting a fuel pin cluster within a nuclear reactor fuel assembly wrapper

    International Nuclear Information System (INIS)

    Marmonier, P.; Mesnage, B.; Teulon, J.; Vayra, J.; Venobre, H.

    1976-01-01

    A supporting member for an array of parallel rails each carrying one row of slidably mounted pins of a fuel cluster is placed coaxially at the lower end of a vertical fuel assembly wrapper. Each parallel rail is provided at each end with a downward extension and terminal lug which engages in a lateral groove formed in the periphery of the supporting member in order to lock and maintain the rails and the fuel pins in uniformly spaced relation within the fuel assembly wrapper. 10 claims, 8 figures

  7. Cell for receipting and dismantling nuclear fuel assembly

    International Nuclear Information System (INIS)

    Beneck, J.A.; Quayre, C.

    1989-01-01

    The cell has a vertical structure with a right section corresponding at that of the assembly to receive, a mechanism for keeping fuel pins at their nominal separation in the form of at least two combs and mechanisms of holding grids and bottom nozzle. The comb arrangements are moved into position by hydraulic actuators so that they cross each other to form a lattice round the fuel pins. The mechanism for holding grid assemblies consist of joints that articulate from a free position to a position where the joints press of the grid on all sides [fr

  8. Spacer grid with mixing blades for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Noailly, J.

    1986-01-01

    The spacer grid for nuclear fuel assembly has two sets of intersecting metal plates provided with blades and defining cells. The plates are fitted only with half-blades associated with a single grid opening. The half-blades of adjacent cells are arranged at 90deg C to each other and each plate has at most one half-blade at each corner of a cell. The invention concerns fuel assemblies of pressurized water reactors. The blades arranged on a single side of the plate provide a good hydraulic uniformity. The invention provides a uniform distribution of blades (and thus of absorbing material in each hydraulic cell) [fr

  9. Fuel assembly duct cutting in the FFTF/IEM Cell

    International Nuclear Information System (INIS)

    Gibbons, P.W.

    1985-01-01

    Two mill type slitting cutters are used in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell during the disassembly sequence of a Driver Fuel Assembly. This disassembly is necessary so that selected parts may be examined both in the IEM Cell and elsewhere. The cutters have been in use for two years. During this time eight Driver Fuel assemblies have been taken apart in the IEM Cell. The cutters' operating philosophy and characteristics, as well as lessons learned from a significant equipment failure are presented. 1 ref., 6 figs., 1 tab

  10. A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

    International Nuclear Information System (INIS)

    Saad, M.; Broeskamp, M.; Hahn, H.; Bignan, G.; Boisset, M.; Silie, P.

    1995-01-01

    The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

  11. 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.

  12. Process for assembling a nuclear fuel element

    International Nuclear Information System (INIS)

    Wachtendonk, H.J. von.

    1984-01-01

    Before insertion into the spacers, the fuel rocks are coated with a self-hardening layer of water-soluble polyvinyl and/or polyether polymer to prevent scratches on the cladding tubes. After insertion, the protective conting is removed by means of water. (orig.) [de

  13. FFTF fuel assembly outlet temperature measurements and comparison to predictions

    International Nuclear Information System (INIS)

    Cramer, E.R.

    1984-06-01

    The data from the FFTF core outlet thermocouples have been valuable in verifying the performance of the core assemblies. The data have been useful to the experimental program and as an aid in understanding some reactor operating phenomena. The thermocouple reliability and repeatability have been good. Almost all of the fueled positions in the core have 3 operable thermocouples and every position has at least one. Differences between the measured assembly outlet temperatures and the calculated outlet temperatures have generally been small. Where significant differences have occurred, an explanation has been found. The difference between the measured and calculated outlet temperatures for each assembly remains constant during the cycle

  14. Mapping device and process for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Vere, B.; Mathevon, P.

    1986-01-01

    The nuclear fuel assembly is mapped while it is loaded. The present invention proposes a simple and sure process of identification of the rods of the assembly, allowing to keep for each one all the information concerning it. The rods are installed as a regular grid in a rack by placing the rods one by one in a former, reading the label of each rod and recording its coordinates before removing the label, aligning the rack with the former, and drawing the rods from the former into the rack. A device for mapping the assembly is also claimed. The invention can be applied to a water cooled and moderated reactor [fr

  15. Design package for fuel retrieval system fuel handling tool modification

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    1998-01-01

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

  16. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  17. A parametric study of assembly pressure, thermal expansion, and membrane swelling in PEM fuel cells

    OpenAIRE

    Maher A.R. Sadiq Al-Baghdadi

    2016-01-01

    Proton Exchange membrane (PEM) fuel cells are still undergoing intense development, and the combination of new and optimized materials, improved product development, novel architectures, more efficient transport processes, and design optimization and integration are expected to lead to major gains in performance, efficiency, durability, reliability, manufacturability and cost-effectiveness. PEM fuel cell assembly pressure is known to cause large strains in the cell components. All components ...

  18. Fuel cycles of WWER-1000 based on assemblies with increased fuel mass

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlovichev, A.; Shcherenko, A.

    2011-01-01

    Modern WWER-1000 fuel cycles are based on FAs with the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively. The highest possible fuel enrichment has reached its license limit that is 4.95 %. Research in the field of modernization, safety justification and licensing of equipment for fuel manufacture, storage and transportation are required for further fuel enrichment increase (above 5 %). So in the nearest future an improvement of technical and economic characteristics of fuel cycles is possible if assembly fuel mass is increased. The available technology of the cladding thinning makes it possible. If the fuel rod outer diameter is constant and the clad inner diameter is increased to 7.93 mm, the diameter of the fuel pellet can be increased to 7.8 mm. So the suppression of the pellet central hole allows increasing assembly fuel weight by about 8 %. In this paper we analyze how technical and economic characteristics of WWER-1000 fuel cycle change when an advanced FA is applied instead of standard one. Comparison is made between FAs with equal time interval between refueling. This method of comparison makes it possible to eliminate the parameters that constitute the operation component of electricity generation cost, taking into account only the following technical and economic characteristics: 1)cycle length; 2) average burnup of spent FAs; 3) specific natural uranium consumption; 4)specific quantity of separative work units; 5) specific enriched uranium consumption; 6) specific assembly consumption. Collected data allow estimating the efficiency of assembly fuel weight increase and verifying fuel cycle characteristics that may be obtained in the advanced FAs. (authors)

  19. Fluid-Structure Interaction in a 3-by-3 Reduced-Scale Fuel Assembly Network

    Directory of Open Access Journals (Sweden)

    Guillaume Ricciardi

    2010-01-01

    network of 3 by 3, subjected to an axial flow. The objective is to analyse the fluid force induced by the motion of the central fuel assembly on the others fuel assemblies. The displacement of the central fuel assembly is imposed, while the others are fixed. Fluid forces acting on fuel assemblies are measured with force sensors. We observed that the coupling between fuel assemblies increases with the fluid velocity, and that the coupling in the transverse direction is not negligible compared to the coupling in the direction of excitation. We also observe that the fluid flow induces a stiffening of the central fuel assembly.

  20. Device for the discharge of fuel assemblies from the core of a reactor cooled with liquid metal to a fuel assembly store

    International Nuclear Information System (INIS)

    Timofeev, A.V.; Batjukov, V.I.; Fadeev, A.I.; Sapkin, A.F.; Scijan, C.G.; Ordynskij, G.V.; Dracev, V.P.; Pogodin, E.N.

    1977-01-01

    In a hermetic chamber, there is a fuel-charging machine with a fuel assembly grab head. It can be moved in channels both to the reactor and to the fuel assembly store. Coolant dripping from fuel assemblies is collected in a container that stretches over the whole distance between the two channels. It consists of number of carriages linked together that may be rolled out of the chamber through a transfer canal in order to take out the collected coolant. (DG) [de

  1. Study of fuel assemblies for the nuclear reactor GFR

    International Nuclear Information System (INIS)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L.

    2008-01-01

    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)

  2. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  3. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    Karvinen, E.

    1981-06-01

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  4. Final Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of a X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as buckling and spring test specimens of cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of neutron fluence monitors installed in the capsule. A new 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 for 95.19 days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 230 {approx} 420 .deg. C. The specimens were irradiated up to a maximum fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.21 {approx} 1.97. The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell and the obtained results will be very valuable for the related researches of the users.

  5. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

  6. The necessity of improvement for the current LWR fuel assembly homogenization method

    International Nuclear Information System (INIS)

    Tang Chuntao; Huang Hao; Zhang Shaohong

    2007-01-01

    When the modern LWR core analysis method is used to do core nuclear design and in-core fuel management calculation, how to accurately obtain the fuel assembly homogenized parameters is a crucial issue. In this paper, taking the NEA C5G7-MOX benchmark problem as a severe test problem, which involves low-enriched uranium assemblies interspersed with MOX assemblies, we have re-examined the applicability of the two major assumptions of the modern equivalence theory for fuel assembly homoge- nization, i.e. the isolated assembly spatial spectrum assumption and the condensed two- group representation assumption. Numerical results have demonstrated that for LWR cores with strong spectrum interaction, both of these two assumptions are no longer applicable and the improvement for the homogenization method is necessary, the current two-group representation should be improved by the multigroup representation and the current reflective assembly boundary condition should be improved by the 'real' assembly boundary condition. This is a research project supported by National Natural Science Foundation of China (10605016). (authors)

  7. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Anniel, Marc; Bordy, Michel-Aristide

    2009-01-01

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  8. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  9. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  10. The procedure of computational evaluation of the margin to CHF in new generation WWER fuel assemblies

    International Nuclear Information System (INIS)

    Loshchinin, V.

    2013-01-01

    A modified and upgraded empirical procedure of the Institute for Physics and Power Engineering (SSC RF-IPPE) has been presented, applicable to the grids-enhancers (EG) of different types located at random over the length of fuel assembly (FA), which allows its application for optimizing the FA and EG designs. (author)

  11. Assemblies and fuel pin behaviour under irradiation in FBR-350

    International Nuclear Information System (INIS)

    Karaulov, V.N.; Blynski, A.P.; Yakovlev, I.L.; Kononova, E.V.

    1998-01-01

    The efficiency of all types of assemblies and fuel pins of the reactor BN-350 is considered in detail. The factors limiting the efficiency are indicated. The behaviour of assemblies with stainless steel ducts is studied. It is shown that the efficiency is restricted in this case by shape and dimensional changes of hexagonal ducts due to radiation swelling and radiation creep of structural materials. The problem of dimensional changes of ducts was solved for cores of reactors BN-350 and BN-600 after testing in BN-350 of experimental assemblies with ducts made of ferritic-martensitic steel 12Cr13Mo2NbVB. For fuel pins of the second type of loading with clad made of stainless steel 0Cr16Ni15Mo3Nb the efficiency is limited by burnup of 13% h.a. and damage dose of 90 dpa. To increase the burnup the core of the BN-350 is supplied by assemblies of modernized type with pins having a large gas plenum and clad made of the stainless steel 0Cr16Ni15Mo2Mn2TiVB that has a good resistance to irradiation. The efficiency of fuel pins of modernized assemblies in the reactor BN-350 core conditions is provided up to 15% h.a. and damage dose 105 dpa. (author)

  12. Analysis of the equalizing holes resistance in fuel assembly spike for lead-based reactor

    International Nuclear Information System (INIS)

    Zhang, Guangyu; Jin, Ming; Wang, Jianye; Song, Yong

    2017-01-01

    Highlights: • A RELAP5 model for a 10 MWth lead-based reactor was built to study the hydrodynamic characteristics between the equalizing holes in the fuel assembly spike. • Different fuel assembly total blockage scenarios and different resistances for different fuel assemblies were examined. • The inherent safety characteristics of the lead-based reactor was improved by optimizing the configuration of equalizing holes in the fuel assembly spike. - Abstract: To avoid the damage of the fuel rod cladding when a fuel assembly (FA) is totally blocked, a special configuration of the fuel assembly spike was designed with some equalizing holes in the center region which can let the coolant to flow during the totally blockage scenarios of FA. To study the hydrodynamic characteristics between the equalizing holes and an appropriate resistance, a RELAP5 model was developed for a 10 MWth lead-based reactor which used lead-bismuth as coolant. Several FA total blockage and partial core blockage scenarios were selected. The simulation results indicated that when all the FA spike equalizing holes had the same hydraulic resistance, only a narrow range of suitable equalizing holes resistances could be chosen when a FA was blocked. However, in the two or more FA blockage scenarios, there were no appropriate resistances to meet the requirement. In addition, with different FA spike equalizing holes with different resistances, a large range of suitable equalizing hole resistances could be chosen. Especially a series of suitable resistances were selected when the small power FA resistance was 1/2, 1/4, 1/8 of the large one. Under these circumstances, one, two or three FA blockages would not damage the core. These demonstrated that selecting a series of suitable hydraulic resistances for the equalizing holes could improve the safety characteristics of the reactor effectively.

  13. Increased local corrosion of SVEA-96 fuel assemblies in KKL. Final report

    International Nuclear Information System (INIS)

    2001-11-01

    In February 1997 it was noted that the cladding surface below the spacers of SVEA-96 fuel assemblies showed increased local corrosion. This phenomenon called 'Enhanced Spacer Shadow Corrosion' (ESSC) by the fuel supplier was carefully monitored during the following 4 years. Several measures have been taken in order to counteract this ESSC. In this report of the Swiss Federal Agency for the Safety of Nuclear Installations (HSK) a summary is given of the technical and licensing aspects of ESSC. Although the fundamental mechanisms for the occurrence of ESSC are not yet sufficiently understood, short-term modification to water chemistry and the increasing use of improved cladding materials have effectively reduced this phenomenon. For the justification of the use of ESSC-damaged SVEA-96 fuel assemblies, HSK established temporary criteria which are based on technical investigations by the fuel assembly supplier. Among these, a special mention can be made of the more restrictive thermo-mechanical operation limit (TMOL) curve. As proof with respect of the HSK criteria, the plant operator conducted extended inspections on fuel assemblies during serving periods in 1997-2001 (measurement of oxide thickness). The conservative aspect of the measurements was assured through destructive examinations carried out at the Hot Laboratory of the Paul Scherrer Institute (PSI). Based on the modified water chemistry and the design of the core loading for cycle 18 (2001/2002) which contains only ESSC resistant cladding materials (LK2+, LK3), the original licence basis concerning the tolerable oxide thickness on the cladding could be guarantied. This has been verified by the results of a fuel assembly examination in August 2001. Therefore, the problem of the increased corrosion of the cladding of the SVEA-96 fuel assemblies is considered as being solved

  14. Studies of thermal-hydraulic flow stability characteristics of two design versions of the SVEA BWR fuel assembly, based on signal recordings made in the Forsmark 2 plant over the years 1990-1995

    International Nuclear Information System (INIS)

    Blomstrand, J.; Carlsson, R.; Petersson, M.

    1997-01-01

    Boiling Water Reactor (BWR) Stability normally refers to a specific topic denoted ''Core Stability''. This phenomenon embraces the entire core, involving periodic fluctuations of a coherent nature in the flows through the core coolant channels. The subject is well known and has been discussed in a large number of conference papers and articles. Other periodic fluctuations which are incoherent may also be present in the channel flows through BWR cores. The phenomenon, denoted ''Channel Flow Stability'', is also well known. Experimental research work on this topic has mostly been made out-of-pile. This paper presents studies of ''Channel Flow Stability'' that have been conducted differently. They are based on the dynamic behaviour of measured channel inlet flows to the core of an operating BWR: Forsmark 2 in Sweden. The investigations cover the years 1990-1995, a time period during which its core inventory experienced a transition from fuel of one assembly design to another, having an increased number of fuel rods of reduced diameter. (UK)

  15. Tritium Systems Test Assembly: design for major device fabrication review

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sherman, R.H.

    1977-06-01

    This document has been prepared for the Major Device Fabrication Review for the Tritium Systems Test Assembly (TSTA). The TSTA is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for fusion reactor systems. The principal objectives for TSTA are: (a) demonstrate the fuel cycle for fusion reactor systems; (b) develop test and qualify equipment for tritium service in the fusion program; (c) develop and test environmental and personnel protective systems; (d) evaluate long-term reliability of components; (e) demonstrate long-term safe handling of tritium with no major releases or incidents; and (f) investigate and evaluate the response of the fuel cycle and environmental packages to normal, off-normal, and emergency situations. This document presents the current status of a conceptual design and cost estimate for TSTA. The total cost to design, construct, and operate TSTA through FY-1981 is estimated to be approximately $12.2 M

  16. Measurements of decay heat and gamma-ray intensity of spent LWR fuel assemblies

    International Nuclear Information System (INIS)

    Vogt, J.; Agrenius, L.; Jansson, P.; Baecklin, A.; Haakansson, A.; Jacobsson, S.

    1999-01-01

    Calorimetric measurements of the decay heat of a number of BWR and PWR fuel assemblies have been performed in the pools at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, CLAB. Gamma-ray measurements, using high-resolution gamma-ray spectroscopy (HRGS), have been carried out on the same fuel assemblies in order to test if it is possible to find a simple and accurate correlation between the 137 CS -intensity and the decay heat for fuel with a cooling time longer than 10-12 years. The results up to now are very promising and may ultimately lead to a qualified method for quick and accurate determination of the decay heat of old fuel by gamma-ray measurements. By means of the gamma spectrum the operator declared data on burnup, cooling time and initial enrichment can be verified as well. CLAB provides a unique opportunity in the world to follow up the decay heat of individual fuel assemblies during several decades to come. The results will be applicable for design and operation of facilities for wet and dry interim storage and subsequent encapsulation for final disposal of the fuel. (author)

  17. Development of a Hi per PWR Fuel assembly for OPR1000 and APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jung Min; Park, Nam Gyu; Jeon, Kyeong Lak; Hwang, Sun Tack [KEPCO Nuclear Fuel Co. Ltd, Daejeon (Korea, Republic of)

    2012-03-15

    KEPCO Nuclear Fuel Co., Ltd. (KNF) launched a new advanced high performance fuel development project in September of 2005, and the HIPER (High Performance with Efficiency and Reliability) fuel was successfully developed in 2010, accomplishing a series of out-of-pile tests. The HIPER Lead Test Assemblies (LTAs) were fabricated in early 2011, and LTA in-reactor verification testing was started in the mid of 2011. The LTAs are scheduled to undergo in-reactor verification testing over four years approximately, and then HIPER fuel will be supplied commercially after obtaining the commercial supply license. In this paper, the development objectives, main design features, including PWR fuel technology development strategy in Korea, out-of-pile test outlines and in-pile test plan for HIPER fuel are described.

  18. Aircraft-Fuel-Tank Design for Liquid Hydrogen

    Science.gov (United States)

    Reynolds, T W

    1955-01-01

    Some of the considerations involved in the design of aircraft fuel tanks for liquid hydrogen are discussed herein. Several of the physical properties of metals and thermal insulators in the temperature range from ambient to liquid-hydrogen temperatures are assembled. Calculations based on these properties indicate that it is possible to build a large-size liquid-hydrogen fuel tank which (1) will weigh less then 15 percent of the fuel weight, (2) will have a hydrogen vaporization rate less than 30 percent of the cruise fuel-flow rate, and (3) can be held in a stand-by condition and readied for flight in a short time.

  19. A Hold-down Margin Assessment using Statistical Method for the PWR Fuel Assembly

    International Nuclear Information System (INIS)

    Jeon, S. Y.; Park, N. K.; Lee, K. S.; Kim, H. K.

    2007-01-01

    The hold-down springs provide an acceptable hold down force against hydraulic uplift force absorbing the length change of the fuel assembly relative to the space between the upper and lower core plates in PWR. These length changes are mainly due to the thermal expansion, irradiation growth and creep down of the fuel assemblies. There are two kinds of hold-down springs depending on the different design concept of the reactor internals of the PWR in Korea, one is a leaf-type hold down spring for Westinghouse type plants and the other is a coil-type hold-down spring for OPR1000 (Optimized Power Reactor 1000). There are four sets of hold-down springs in each fuel assembly for leaf type hold-down spring and each set of the hold-down springs consists of multiple tapered leaves to form a cantilever leaf spring set. The length, width and thickness of the spring leaves are selected to provide the desired spring constant, deflection range, and hold down force. There are four coil springs in each fuel assembly for coil-type hold-down spring. In this study, the hold-down forces and margins were calculated for the leaf-type and coil-type hold-down springs considering geometrical data of the fuel assembly and its components, length changes of the fuel assembly due to thermal expansion, irradiation growth, creep, and irradiation relaxation. The hold-down spring forces were calculated deterministically and statistically to investigate the benefit of the statistical calculation method in view of hold-down margin. The Monte-Carlo simulation method was used for the statistical hold down force calculation

  20. 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)

  1. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    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, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of 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

  2. A mathematical model of an automatic assembler to stack fuel pellets

    International Nuclear Information System (INIS)

    Jarvis, R.G.; Joynes, R.; Bretzlaff, C.I.

    1980-11-01

    Fuel elements for CANDU reactors are assembled from stacks of cylindrical UO 2 pellets, with close tolerances on lengths and diameters. Present stacking techniques involve extensive manual operations and they can be speeded up and reduced in cost by an automated device. If gamma-active fuel is handled such a device is essential. An automatic fuel pellet assembly process was modelled mathematically. The model indicated a suitable sequence of pellet manipulations to arrive at a stack length that was always within tolerance. This sequence was used as the inital input for the design of mechanical hardware. The mechanical design and the refinement of the mathematical model proceeded simultaneously. Mechanical constraints were allowed for in the model, and its optimized sequence of operations was incorporated in a microcomputer program to control the mechanical hardware. (auth)

  3. 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)

  4. 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.

  5. Effect of Heterogeneity of JSFR Fuel Assemblies to Power Distribution

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Shimazu, Yoichiro; Hibi, Koki; Fujimura, Koji

    2013-01-01

    Conclusion: 1) Strong heterogeneity of JSFR assemblies was successfully calculated by BACH. 2) Verification test of BACH: • Infinite assembly model; • Color set model; • Good agreement with Monte-Carlo results. 3) Core calculations 3 models for inner duct was used; inward model, outward model and homogeneous model. • k eff difference between the inward and out ward model → 0.3%Δk; • ~20% effect on flux and power distributions. Therefore, we have to pay careful attention for the location of inner duct in fuel loading of JSFR

  6. Premixer assembly for mixing air and fuel for combustion

    Science.gov (United States)

    York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

    2016-12-13

    A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

  7. Fail-safe storage rack for irradiated fuel rod assemblies

    Science.gov (United States)

    Lewis, D.R.

    1993-03-23

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  8. MITG test assembly design and fabrication

    International Nuclear Information System (INIS)

    Schock, A.

    1983-01-01

    The design, analysis, and evaluation of the Modular Isotopic Thermoelectric Generator (MITG), described in an earlier paper, led to a program to build and test prototypical, modules of that generator. Each test module duplicates the thermoelectric converters, thermal insulation, housing and radiator fins of a typical generator slice, and simulates its isotope heat source module by means of an electrical heater encased in a prototypical graphite box. Once the approx. 20-watt MITG module has been developed, it can be assembled in appropriate number to form a generator design yielding the desired power output. The present paper describes the design and fabrication of the MITG test assembly, which confirmed the fabricability of the multicouples and interleaved multifoil insulation called for by the design. Test plans, procedures, instrumentation, results, and post-test analyses, as well as revised designs, fabrication procedures, and performance estimates, are described in subsequent papers in these proceedings

  9. Mechanical characterization tests of the X2-Gen fuel assembly and skeleton

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kang, Heung Seok

    2011-01-01

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the X2-Gen fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the X2-Gen was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  10. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  11. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

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

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari

    1997-01-01

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

  13. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    International Nuclear Information System (INIS)

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  14. Design of a nuclear fuel rod support grid using axiomatic design

    International Nuclear Information System (INIS)

    Song, Kee Nam; Yoon, Kyung Ho; Kang, Byung Soo; Park, Gyung Jin; Choi, Sung Kyoo

    2002-01-01

    Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains a coolable geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design

  15. Design of remote handled process assemblies for the process facility modifications project

    International Nuclear Information System (INIS)

    Smets, J.L.; Ajifu, D.A.

    1987-01-01

    The modular design philosophy for the process facility modification project utilizes an integrated design of components to facilitate operations and maintenance of nuclear fuel reprocessing equipment in a hot cell environment. The utilization of a matrix of remoteable base frames combines with process equipment designed as remote assemblies and sub-assemblies has simplified the overall design. Modularity will allow future flexibility while providing advantages for construction and maintenance in the initial installation

  16. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  17. A comparison of spent fuel assembly control instruments: The Cadarache PYTHON and the Los Alamos Fork

    International Nuclear Information System (INIS)

    Bignan, G.; Capsie, J.; Romeyer-Dherbey, J.

    1991-01-01

    Devices to monitor spent fuel assemblies while stored under water with nondestructive assay methods, have been developed in France and in the United States. Both devices are designed to verify operator's declared values of exposures and cooling-time but the applications and thus the designs of the systems differ. A study, whose results are presented in this paper, has been conducted to compare the features and the performances of the two instruments. 4 refs., 9 figs

  18. Process and device for fabricating nuclear fuel assembly grids

    International Nuclear Information System (INIS)

    Thiebaut, B.; Duthoo, D.; Germanaz, J.J.; Angilbert, B.

    1991-01-01

    The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

  19. A simplified approach to design for assembly

    DEFF Research Database (Denmark)

    Moultrie, James; Maier, Anja

    2014-01-01

    The basic principles of design for assembly (DfA) are well established. This paper presents a short review of the development of DfA approaches before presenting a new tool in which these principles are packaged for use in teams, both in an industrial and an educational context. The fundamental c...... that includes custom designed post-it notes and simple check lists for scoring. The process is demonstrated in a single case study in an Indian firm....

  20. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  1. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

    2015-01-01

    In 2005, six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in South Ukraine Unit 3 (SU3). The LTAs completed the planned four cycles of operation and reached an average assembly burnup in excess of 43 MWd/ kgU. Post Irradiation Examination (PIE) inspections were performed after completion of each cycle and it was concluded that the 6 Westinghouse LTAs performed as expected during their operational regimes. In 2010, a full region of 42 assemblies of an enhanced WWER-1000 fuel design for Ukrainian reactors, designated WFA, was loaded in SU3. The WFA includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. In 2015, 26 WFAs completed their planned four cycles of operation reaching an average assembly burnup in excess of 42 MWd/ kgU. Currently 36 WFAs continue operating their fourth cycle in SU3. In addition, South Ukraine Unit 2 (SU2) has been loaded with WFAs and 27 assemblies have completed two cycles of operation reaching an average assembly burnup above 24 MWd/kgU. PIE for the WFAs has been completed after each cycle of operation. All assemblies have been examined for visible damage or non-standard position of fuel assembly components during unloading and reloading. All WFAs have also been subject to the standard leak testing process, with all fuel rods found to be hermetically sealed and non-leaking. Each outage, six WFAs have been subject to a more extensive inspection program. In 2012, 2013, and 2015, the Westinghouse Fuel Inspection and Repair Equipment (FIRE) workstation were used for the SU3 inspections. Excellent irradiation fuel performance has been observed and measured on all WFAs. The fuel assembly growth, rod cluster control assembly (RCCA) drag forces, oxide thickness, total fuel rod-to-nozzle gap channel closure, and fuel assembly bow data were within the bounds of the Westinghouse experience database. Results and concluding remarks from the PIEs are provided in this paper. In

  2. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    Tarvainen, M.; Baecklin, A.; Haakanson, A.

    1992-02-01

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

  3. Turbulent mixing in the foot piece of a HPLWR fuel assembly

    International Nuclear Information System (INIS)

    Hofmeister, Jan; Laurien, Eckart; Class, Andreas G.; Sculenberg, Thomas

    2005-01-01

    A homogeneous turbulent mixing of coolant flows with different temperatures at the fuel assembly inlets is an important requirement to minimize hot spots in a fuel assembly of a High Performance Light Water Reactor (HPLWR). Therefore, the mixing chamber between lower core plate, flow adjuster and the mixing chamber within the cluster foot piece diffuser have been investigated using the Computational Fluid Dynamics (CFD)-code Fluent 6.1 and its implemented k-ε model. The previously presented 3D-CAD-geometry has been simplified using Gambit 2.1.2 and consists of various inlet and outlet tubes or channels in the foot piece bottom plate, the lower core plate and the flow adjuster establishing the boundaries of two consecutive mixing chambers. The temperature distribution at the inlet of the sub-channels of the cluster fuel assemblies is presented. It reveals temperature variations at the coolant inlet of the nine fuel assemblies which are not acceptable. Therefore, a design modification to improve mixing is developed, thus reducing the temperature variation below 3K. The accuracy of the simulations has been studied and quantified. (author)

  4. Investigation of IFMIF target assembly structure design

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Sugimoto, Masayoshi; Yamamura, Toshio

    2006-10-01

    In the International Fusion Materials Irradiation Facility (IFMIF), the back-wall of target assembly is the part suffered the highest neutron-flux. The back-wall and the assembly are designed to have lips for cutting/welding at the back-wall replacement. To reduce thermal stress and deformation of the back-wall under neutron irradiation, contact pressure between the back-wall and the assembly is one of dominant factors. Therefore, an investigation was performed for feasible clamping pressure of a mechanical clamp set in limited space around the back-wall. It was clarified that the clamp can give a pressure difference up to 0.4 MPa between the contact pressure and atmosphere pressure in the test cell room. Also a research was performed for the dissimilar metal welding in the back-wall. Use of 309 steel was found adequate as the intermediate filler metal through the research of previous welding. Maintaining a temperature of the target assembly so as to avoid a freezing of liquid lithium is needed at the lithium charge into the loop before the beam injection. The assembly is covered with thermal insulation. Therefore, a research and an investigation were performed for compact and light thermal-insulation effective even under helium (i.e. high heat-conduction) condition of the test cell room. The result was as follows; in the case that a thermal conductivity 0.008 W/m·K of one of found insulation materials is available in the temperature range up to 300degC of the IFMIF target assembly, needed thickness and weight of the insulation were respectively only 8.2 mm and 32 kg. Also a research was performed for high-heat-density heaters to maintain temperature of the back-wall which can not be cover with insulation due to limited space. A heater made of silicon-nitride was found to be adequate. Total heat of 8.4 kW on the back-wall was found to be achievable through an investigations of heater arrange. Also an investigation was performed for remote-handling device to

  5. Design, Development and Installation of Jordan Subcritical Assembly

    Directory of Open Access Journals (Sweden)

    Ned Xoubi

    2013-01-01

    Full Text Available Following its announcement in 2007 to pursue a nuclear power program and in the absence of any nuclear facility essential for the education, training, and research, Jordan decided to build a subcritical reactor as its first nuclear facility. Jordan Subcritical Assembly (JSA is uranium fueled light water moderated and reflected subcritical reactor driven by a plutonium-beryllium source, and the core consists of 313 LEU fuel rods, loaded into a water-filled vessel in a square lattice of 19.11 mm pitch. The fuel rods are based on PWR fuel structural pattern type, made of uranium oxide (UO2 with 3.4 wt% 235U enrichment in zirconium alloy (Zr-4 cladding. Design, optimization, and verification were performed using MCNP5 nuclear code; the computed effective multiplication factor is 0.95923. The JSA is designed to fulfill the training needs of students and is equipped to perform all of the fundamental experiments required for a typical nuclear engineering university program. This paper presents the design, development, modeling, core analysis, and utilization of Jordan’s first nuclear facility and why this simplified low cost reactor presents an attractive choice to fulfill the preliminary experimental needs of nuclear engineering education in developing countries.

  6. Validation of structural design of JHR fuel element

    International Nuclear Information System (INIS)

    Brisson, S.; Miras, G.; Le Bourdonnec, L.; Lemoine, P.; Anselmet, M.C.; Marelle, V.

    2010-01-01

    The validation of the structural design of the Jules Horowitz Reactor fuel element was made by the Finite Element Method, starting from the Computer Aided Design. The JHR fuel element is a cylindrical assembly of three sectors composed of eight rolled fuel plates. A roll-swaging process is used to join the fuel plates to three aluminium stiffeners. The hydraulic gap between each plate is 1.95 mm. The JHR fuel assembly is fastened at both ends to the upper and lower endfittings by riveting. The main stresses are essentially thermal loads, imposed on the fuel zone of the plates. These thermal loads result from the nuclear heat flux (W/cm 2 ). The mechanical loads are mainly hydraulic thrust forces. The average coolant velocity is 15 m/s. Seismic effects are also studied. The fuel assembly is entirely modelled by thin shells. The model takes into account asymmetric thermal loads which often appear in Research Reactors. The mechanics of the fuel plates vary in function of the burn up. These mechanical properties are derived from the data sets used in the MAIA code, and the validity of the structure is demonstrable at throughout the life of the fuel. Results concerning displacement are compared to functional criteria, while results concerning stress are compared to RCC-MX criteria. The results of this analysis show that the mechanical and geometrical integrity of the JHR fuel elements is respected for Operating Categories 1 and 2. This paper presents the methodology of this demonstration for the results obtained. (author)

  7. Seismic analysis of fuel and target assemblies at a production reactor

    International Nuclear Information System (INIS)

    Braverman, J.I.; Wang, Y.K.

    1991-01-01

    This paper describes the unique modeling and analysis considerations used to assess the seismic adequacy of the fuel and target assemblies in a production reactor at Savannah River Site. This confirmatory analysis was necessary to provide assurance that the reactor can operate safely during a seismic event and be brought to a safe shutdown condition. The plant which was originally designed in the 1950's required to be assessed to more current seismic criteria. The design of the reactor internals and the magnitude of the structural responses enabled the use of a linear elastic dynamic analysis. A seismic analysis was performed using a finite element model consisting of the fuel and target assemblies, reactor tank, and a portion of the concrete structure supporting the reactor tank. The effects of submergence of the fuel and target assemblies in the water contained within the reactor tank can have a significant effect on their seismic response. Thus, the model included hydrodynamic fluid coupling effects between the assemblies and the reactor tank. Fluid coupling mass terms were based on formulations for solid bodies immersed in incompressible and frictionless fluids. The potential effects of gap conditions were also assessed in this evaluation. 5 refs., 6 figs., 1 tab

  8. 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.

  9. Investigation regarding the safety of handling the fuel assemblies for the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    It was concluded previously that the general inspection of safety and the repair of shielding can be carried out as the fuel assemblies are charged, and the safety can be secured sufficiently. According to the decision by the meeting of cabinet ministers concerned with the nuclear ship ''Mutsu'', the Mutsu General Inspection and Repair Technology Investigation Committee investigated on the basic concept regarding the method and the safety of taking out, transporting and preserving the fuel assemblies. 112 fuel rods and 9 burnable poison rods are arranged into the square grid of 11 x 11 in a fuel assembly, and 32 fuel assemblies are employed. The contents of the investigation are the outline of the fuel assemblies, the present states of nuclear fission products, surface dose rate and soundness of the fuel assemblies, the safety of taking out, transporting and preserving the fuel assemblies, the measures required for securing the safety, and the place for taking out the fuel assemblies. In case of taking out, transporting and preserving the fuel assemblies, it is considered in view of the present state of the fuel assemblies that the safety can be secured sufficiently if the works are carried out carefully by taking the methods and conditions investigated into consideration. Also the committee reached already the conclusion described at the outset. (Kako, I.)

  10. Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

    2006-01-01

    According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

  11. Core-control assembly with a fixed fuel support

    International Nuclear Information System (INIS)

    Challberg, R.C.

    1993-01-01

    A core-control assembly is described comprising: a control rod having a plurality of blades; a control-rod guide tube for guiding vertical motion of said control rod; a fuel support for supporting fuel bundles separated by said blades, said fuel support having an aperture conforming to a cross section of said control rod through said blades for preventing rotational movement of said control rod to a decoupling orientation when said control rod is between a maximum power position and a minimum power position, said minimum power position being above said maximum power position, said fuel support being supported by said control-rod guide tube; control-rod drive means for controlling vertical motion of said control rod, said control-rod drive means providing for vertical motion between said maximum power position and said minimum power position, said control-rod drive means providing for vertical movement to a decoupling position, said decoupling position being no lower than said minimum power position, said decoupling position being at a level sufficient to permit said control rod to rotate to a decoupling orientation relative to said fuel support; and coupling means for coupling said control rod to said control rod drive means, said coupling means being releasable by rotational movement of said control rod to said decoupling orientation relative to said control-rod drive means

  12. Dynamic responses of fuel and target assemblies of a production reactor

    International Nuclear Information System (INIS)

    Crowley, D.A.; Yau, W.F.

    1983-01-01

    As part of the qualification research aimed at assuring safe operation of the production reactors at the Savannah River Plant (SRP), the dynamic reponses of internal reactor components are being analyzed. One such program investigates the responses of heavy fuel and target assemblies undergoing two types of loading - the distrurbances due to the motion of machines that transport the assemblies to and from the reactor, and the seismic loading due to a design basis earthquake during reactor operation. This qualification research is supported by an experimental program to verify the analytical predictions

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

    International Nuclear Information System (INIS)

    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

  14. 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

  15. Results of Parametric Design Studies of MOX Lead Test Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.; Bychkov, S.A.; Lazarenko, A.A.; Sidorenko, V.D.; Styrin, Y.A.

    1998-12-01

    The parametric studies of MOX LTA design have been executed to choose plutonium content in assembly zones for two options of MOX LTA: 3-zones and Island. For 3-zones (100% Plutonium) MOX LTA the fissile plutonium content composition of 4.2%/3,0%/2% has been chosen. MOX LTA of the chosen compositions has been studied by using multi-assembly configuration that allows investigating of influence of MOX LTA environment: uranium assemblies of different irradiation. Plutonium Island with 54 plutonium pins in the center of MOX LTA has been considered in two modifications: uniform island; and graded island with lower plutonium content in one peripheral row of pins. It is shown that plutonium content in the uniform island cannot exceed 2.7% because of adopted power peaking limitations and therefore this design seems unreasonable for practical use. For graded island the plutonium content composition 3.8%/2.8% with uranium environment of 3.7% U-235 has been chosen. Evolution of assembly power and burnup distributions, inter-pin power and isotopic distributions while fuel irradiating have been analyzed. In addition to the base uranium environment of 3.7%, a set of calculations has been executed for 4.4%. Most of the studies have been executed by the code TVS-M that is at the final stage of licensing and it is to be used in the nearest future as a base instrument for VVER core calculations while using both uranium and MOX fuel. So the obtained results must be considered as preliminary ones and they demand additional analysis and investigations.

  16. Results of Parametric Design Studies of MOX Lead Test Assembly

    International Nuclear Information System (INIS)

    Pavlovitchev, A.M.; Bychkov, S.A.; Lazarenko, A.A.; Sidorenko, V.D.; Styrin, Y.A.

    1998-01-01

    The parametric studies of MOX LTA design have been executed to choose plutonium content in assembly zones for two options of MOX LTA: 3-zones and Island. For 3-zones (100% Plutonium) MOX LTA the fissile plutonium content composition of 4.2%/3,0%/2% has been chosen. MOX LTA of the chosen compositions has been studied by using multi-assembly configuration that allows investigating of influence of MOX LTA environment: uranium assemblies of different irradiation. Plutonium Island with 54 plutonium pins in the center of MOX LTA has been considered in two modifications: uniform island; and graded island with lower plutonium content in one peripheral row of pins. It is shown that plutonium content in the uniform island cannot exceed 2.7% because of adopted power peaking limitations and therefore this design seems unreasonable for practical use. For graded island the plutonium content composition 3.8%/2.8% with uranium environment of 3.7% U-235 has been chosen. Evolution of assembly power and burnup distributions, inter-pin power and isotopic distributions while fuel irradiating have been analyzed. In addition to the base uranium environment of 3.7%, a set of calculations has been executed for 4.4%. Most of the studies have been executed by the code TVS-M that is at the final stage of licensing and it is to be used in the nearest future as a base instrument for VVER core calculations while using both uranium and MOX fuel. So the obtained results must be considered as preliminary ones and they demand additional analysis and investigations

  17. Nuclear reactor having thermally compensated support structure for a fuel assembly

    International Nuclear Information System (INIS)

    Borst, R.

    1980-01-01

    A thermal expansion compensation system for nuclear reactor fuel assemblies is disclosed which utilizes materials with different rates of thermal expansion in appropriate components so to: retain alignment of the assembly; reduce or eliminate thermal bow; and reduce or eliminate jump movement of fuel assemblies. (orig.)

  18. Hierarchy level scheme for quasi-optimum fuel assembly loading in boiling water reactors

    International Nuclear Information System (INIS)

    Sekimizu, K.

    1978-01-01

    A quasi-optimum fuel assembly allocation scheme for boiling water reactors was proposed and confirmed. It is characteristic of the scheme that the criteria function is represented by fuel assembly allotment to fuel groups. For each fuel group, a required property is given beforehand, and fuel assemblies are allocated to the core to determine the group property as closely as possible. By using the scheme, a fuel assembly allocation is obtained that has a large cycle burnup within a restriction for the peak-to-average power ratio. Another allocation is obtained that results in a large burnup of discharged fuel using a different criteria function. However, it is impossible to obtain a strictly optimum solution for a given criteria function because of the vast number of possible fuel assembly allocations. The search range is reduced by adopting a two-step scheme. In the first step, an optimum allocation of fresh assemblies is searched for, based on proper criteria. Then, in the second step, without moving the fresh fuel assemblies, an allocation of reload fuel assemblies is determined that ascertains the required group property as closely as possible. Results of the numerical calculation show that the scheme is very useful for practical fuel assembly allocation

  19. Machine to compact fuel assemblies and to shear out the end parts

    International Nuclear Information System (INIS)

    Auchapt, P.; Sablier, R.; Symard, J.; Seyfried, P.

    1983-01-01

    The present machine allows a simultaneous compaction of the fuel assemblies and the separation of their end parts; the machine compacts the protection structures of the fuel without deforming them and ensures the evacuation of the end parts out of the shears, what allows to consider them apart from α wastes. The present machine accepts fuel assemblies of different length [fr

  20. Control component structure and its removal from fuel assembly

    International Nuclear Information System (INIS)

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

    1982-01-01

    This invention provides methods and apparatus for securing and removing burnable poison rods to the spider in a fuel assembly. A pin is secured to one of the transverse ends of a burnable poison rod. The pin is seated in a bore that is formed in the spider arm appropriate to the rod under consideration. The burnable poison rod is separated from the spider arm by applying a force in a direction that is coincident with the longitudinal axis of the rod and its associated pin. The force is of sufficient magnitude to press the pin out of the spider arm

  1. Handling apparatus for fuel assemblies in a core

    International Nuclear Information System (INIS)

    Hatakenaka, Hideo.

    1975-01-01

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

  2. Fluid flow plate for decreased density of fuel cell assembly

    Science.gov (United States)

    Vitale, Nicholas G.

    1999-01-01

    A fluid flow plate includes first and second outward faces. Each of the outward faces has a flow channel thereon for carrying respective fluid. At least one of the fluids serves as reactant fluid for a fuel cell of a fuel cell assembly. One or more pockets are formed between the first and second outward faces for decreasing density of the fluid flow plate. A given flow channel can include one or more end sections and an intermediate section. An interposed member can be positioned between the outward faces at an interface between an intermediate section, of one of the outward faces, and an end section, of that outward face. The interposed member can serve to isolate the reactant fluid from the opposing outward face. The intermediate section(s) of flow channel(s) on an outward face are preferably formed as a folded expanse.

  3. Method of assembling spent nuclear fuel storage rack

    International Nuclear Information System (INIS)

    Igarashi, Ryokichi; Hasegawa, Hidenobu.

    1982-01-01

    Purpose: To improve the safety of a spent fuel storage rack by stably installing the spent fuel in a pool without using supporting beams. Constitution: A restricted unit is composed of a plurality of spuare cylinders. A plurality of such restricted units are aligned in a direction perpendicularly to the arraying direction of the cylinders in the respective restricted units, are coupled with long connecting plates, and are fixed by welding on a common small base, thereby forming a restricted body. According to such assembling method, a plurality of restricted bodies are connected in a direction that the respective restricted bodies are readily overturned, and are secured to the common base. Accordingly, the restricted bodies can be stably installed in a pool without using supporting beams as the conventional one. (Sekiya, K.)

  4. Development of image processing software for measurement to fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Sang; Koo, Dae Suo; Min, Duk Ki

    1997-12-01

    The software has been developed for rapid fuel rod measurement with image processing method in PIEF pool. It has many image enhancement filtering algorithms which are simplified for easy usage. And extravagant memory problem which may be commonly generated in image processing program is solved by programming technology. Therefore it`s memory is not go to excess in spite of too many filtering operation. when user point at one point in computer monitor screen for measurement with mouse, the program measures the length and etc of binary image`s screen in center of that point. The measurement to fuel assembly mock-up for test had the result of {+-}0.5 mm measurement error. (author). 4 refs., 2 tabs., 41 figs.

  5. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-01-01

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

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

    International Nuclear Information System (INIS)

    Van Santen, A.; Elofsson, K.

    1975-01-01

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

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

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    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 232Th, 238U and 237Np 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.

  8. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-03-14

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders.

  9. Strobe sequence design for haplotype assembly

    Science.gov (United States)

    2011-01-01

    Background Humans are diploid, carrying two copies of each chromosome, one from each parent. Separating the paternal and maternal chromosomes is an important component of genetic analyses such as determining genetic association, inferring evolutionary scenarios, computing recombination rates, and detecting cis-regulatory events. As the pair of chromosomes are mostly identical to each other, linking together of alleles at heterozygous sites is sufficient to phase, or separate the two chromosomes. In Haplotype Assembly, the linking is done by sequenced fragments that overlap two heterozygous sites. While there has been a lot of research on correcting errors to achieve accurate haplotypes via assembly, relatively little work has been done on designing sequencing experiments to get long haplotypes. Here, we describe the different design parameters that can be adjusted with next generation and upcoming sequencing technologies, and study the impact of design choice on the length of the haplotype. Results We show that a number of parameters influence haplotype length, with the most significant one being the advance length (distance between two fragments of a clone). Given technologies like strobe sequencing that allow for large variations in advance lengths, we design and implement a simulated annealing algorithm to sample a large space of distributions over advance-lengths. Extensive simulations on individual genomic sequences suggest that a non-trivial distribution over advance lengths results a 1-2 order of magnitude improvement in median haplotype length. Conclusions Our results suggest that haplotyping of large, biologically important genomic regions is feasible with current technologies. PMID:21342554

  10. Mixcore safety analysis approach used for introduction of Westinghouse fuel assemblies in Ukraine

    International Nuclear Information System (INIS)

    Abdullayev, A.; Baidullin, V.; Maryochin, A.; Sleptsov, S.; Kulish, G.

    2008-01-01

    Six Westinghouse Lead Test Assemblies (LTA) were installed in 2005 and are currently operated in Unit 3 of the South Ukraine NPP (SUNPP) under the Ukraine Nuclear Fuel Qualification Project. At the early stages of the LTAs implementation in Ukraine, there was no experience of licensing of new fuel types, which explains the need to develop approaches for safety substantiation of LTAs. This presentation considers some approaches for performing of safety analysis of the design basis Initiating Events (IE) for the LTA fuel cycles. These approaches are non-standard in terms of the established practices for obtaining the regulatory authorities' permission for the core operation. The analysis was based on the results of the FA and reactor core thermal hydraulic and nuclear design

  11. Device for absorbing the axial forces occurring on the fuel assemblies during operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Sankovich, M.F.

    1978-01-01

    The fuel assemblies consisting of rod-shaped fuel rods stand on a grid plate. Opposite the projections of the upper grid plate mounted on a support barrel the fuel assemblies are elastically supported in order to compensate the mechanical vibrations and thermal expansions occurring during operation. This is achieved by combined bending and torsion springs bridging the distance between projections and fuel assembly end pieces. The bending and torsion springs consist of a bending arm, a torsion piece, and another bending arm being deflected by 90 0 and provided at the end with an upsetting. Each spring consists of round stock. In order to increase the flexibility one of the bending arms is designed conically or stepped. (DG) [de

  12. Coil-springs used as mechanical filter. Modification of the bottom tie plate of a fuel assembly

    International Nuclear Information System (INIS)

    Nylund, O.

    1993-01-01

    Describes an improved design of the bottom tie plate of a fuel assembly. The improvement of the design is an arrangement of horizontal channels holding coil-springs and crossing the vertical channels for the cooling water. The coil-springs work as strainers for the cooling water

  13. Supplemental information for a notice of construction 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 ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). 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 to be used 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, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  14. 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.

  15. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    The report provides a summary of the design of the canister for final disposal of 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 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  16. 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.

  17. Distribution of fission products in graphite sleeves and blocks of the eleventh and twelfth OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Fukuda, Kousaku; Kikuchi, Teruo; Tsuruta, Harumichi.

    1994-06-01

    The 11th and 12th fuel assemblies were irradiated in an in-pile gas loop, OGL-1, installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Distribution of fission products in the graphite sleeves and blocks of the assemblies was measured by gamma-ray spectrometry. The 11th fuel assembly was aimed at testing the irradiation performance of mass product fuels in trial manufacturing of the first charge fuel for the High Temperature Engineering Test Reactor (HTTR) in relatively short irradiation, and the 12th assembly in long-term irradiation. The 12th assembly attained a burnup approximately as high as that of the HTTR driver fuel design. In the graphite sleeve of the 11th assembly, high concentration peaks of fission products were found in the axial distribution. Exposure of failed fuel particles was not detected on the surface of fuel compacts, while fissures of graphite matrix at overcoat boundaries were observed on the surface. These results led to a presumption that fission products, which were released from failed particles located inside of the fuel compact, was easily transported through the fissures of the matrix to the inner surface of the sleeve. In the graphite sleeve of the 12th assembly, 110m Ag was detected together with other fission products of 137 Cs, 134 Cs etc. Silver-110m showed characteristic distribution: this nuclides was less concentrated at the region with highly concentrated 60 Co which is presumed to have been transported from melted sheath material of thermocouples to the graphite sleeve. It was inferred from the distribution that the transport behavior of 110m Ag had been influenced by co-sorption or by pore structure change in the graphite material of the sleeve, which had been induced by metallic elements including cobalt. (author)

  18. 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.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Nuclear fuel assembly top nozzle with improved arrangement of hold-down leaf spring assemblies

    International Nuclear Information System (INIS)

    De Mario, E.E.; Lawson, C.N.

    1991-01-01

    This patent describes a top nozzle for use in a fuel assembly having guide thimbles for mounting the top nozzle. It comprises: a lower adapter plate having a periphery bounding an interior thereof mountable to the guide thimbles: guide structures attached to and extending along the periphery of the adapter plate and upwardly therefrom; an upper hold-down plate mounted to the guide structures for slidable movement relative thereto such that the upper plate can move toward and away from the interior of the lower plate within the space bounded by the guide structures as the upper plate slidably moves along the guide structures; and leaf spring assemblies interposed between and engaged with the lower and upper plates so as to yieldably support the upper plate in spaced relation above the lower plate and bias the upper plate for movement away from the lower plate; the leaf spring assemblies being provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies cross the interior of the lower plate in a diagonal fashion between adjacent ones of the guide structures

  1. Apparatus for adjusting the elevation of fuel rods in a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Hale, D.L.; Culbreth, T.F.

    1988-01-01

    A tool adapted for adjusting the level of a nuclear fuel rod in a fuel assembly is described comprising: an expander comprising two elongate generally parallel and laterally spaced apart arms extending in a longitudinal direction, an actuator operatively mounted to the expander so as to be disposed between the two arms and being movable with respect thereto, and cooperating surface means mounted to the arms and the actuator for laterally separating the free ends of the arms to a predetermined maximum distance upon movement of the actuator with respect to the arms

  2. Measurements on spent-fuel assemblies at Arkansas Nuclear One using the Fork system. Final report, January 1995

    International Nuclear Information System (INIS)

    Ewing, R.I.; Bronowski, D.R.; Bosler, G.E.; Siebelist, R.; Priore, J.; Hansford, C.H.; Sullivan, S.

    1997-03-01

    The Fork measurement system has been used to examine spent-fuel assemblies at the two reactors of Arkansas Nuclear One, operated by Entergy Operations, Inc. The Unit 1 reactor is a Babcock and Wilcox (B and W) design, and the Unit 2 reactor is a Combustion Engineering (CE) design. The neutron and gamma-ray emissions from individual spent-fuel assemblies were measured in the storage pools by raising each assembly pathway out of the storage rack and performing a measurement near the center of the assembly. The overall accuracy of the measurements after corrections is about 2%. Thirty-four assemblies were examined at Unit 1, and forty-one assemblies at Unit 2. The average deviation of the burnup measurements from the calibration was 3.0% at Unit 1 and 3.5% at Unit 2, indicating 2 to 3% random variation among the reactor records. There was no indication of clearly anomalous assemblies. Axial Scans of the variation in neutron and gamma ray emission were obtained by collecting data at several locations along the length of three assemblies at Unit 2. Two of these assemblies were nonstandard in that each contained a small neutron source. The sources were detected by the axial scans. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Entergy Operations, Inc., the Electric Power Research Institute, and the Office of Civilian Radioactive Waste Management of the US Department of Energy

  3. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    Garcia V, M.A.

    2006-01-01

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

  4. Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

    International Nuclear Information System (INIS)

    Ferreira, W.J.; Luz, M.

    1985-01-01

    The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

  5. Conceptual design report of the SMART fuel rod

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  6. 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.

  7. 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).

  8. Operational experience with the first eighteen slightly enriched uranium fuel assemblies in the Atucha-1 nuclear power plant

    International Nuclear Information System (INIS)

    Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Fink, J.; Casario, J.A.; Alvarez, L.

    1997-01-01

    Atucha I is a 357 Mwe nuclear station, moderated and cooled with heavy water, pressure vessel type of German design, located in Argentina. Fuel assemblies (FA) are 36 active natural UO2 rod clusters, 5.3 meters long and fuel is on power. Average FA exit burnup is 6 MWd/kg U. The reactor core contains 252 FA. To reduce the fuel costs about 6 MU$S/yr a program of utilization of SEU (0.85 %w U235) fuel was started at the beginning of 1995 with the introduction of 12 FA in the first step. The exit burnup of FA is approx. 10 MWd/kgU. It is planned to increase gradually the number of them up to having a full core with SEU fuel with an expected FA average exit burnup of 11 MWd/kgU. The SEU program has also the advantage of a strong reduction of spent fuel volume, and a moderate reduction of fuelling machine use. This paper presents the satisfactory operation experience with the introduction of the first 12 SEU fuel assemblies and the planned activities for the future. The fresh SEU fuel assemblies were introduced in six fuel channels located in an intermediate zone located 136 cm from the center of the reactor and selected because they have higher margins to the channel powers limits to accommodate the initial 15 to 20 % relative channel power increase. To verify the design and fuel management calculations, comparisons have been made of the calculated and measured values of the variation of channel ΔT, regulating rods insertion and flux reading in in-core detectors near to the refueled channel. The agreement was good and in most of the cases within the measurement errors. Cell calculations were made with WIMS-D4, and reactor calculations with PUMA. a fuel management 3D diffusion program developed in Argentina. With SEU fuel with a greater burnup in the central high power core region, new operating procedures were developed to prevent PCI failures in fuel power ramps that arise during operation. Some fuel rod and structural assembly design changes were introduced on the

  9. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  10. Determination of fuel assembly vibrational modes through analysis of incore detector noise

    International Nuclear Information System (INIS)

    Johnson, R.S.

    1986-01-01

    In order to better characterize fuel assembly vibration at Duke Power Company's Oconee Nuclear Station, incore noise data were acquired an analyzed from prompt responding detectors incorporated in the Oconee 2, Cycle 7 core. Duke Power Company began actively pursuing an inhouse Neutron Noise Analysis program for routine surveillance of reactor internals vibration in 1979. Noise data has since been acquired and analyzed for twelve cycles of operation for the three Oconee units. Duke Power's Oconee Unit 2 is a Babcock and Wilcoxs pressurized water reactor with a rate thermal power of 2568MW. For Oconee 2, Cycle 7 operation, two test assemblies, each employing a string of seven axially-spaced, prompt responding hafnium detectors, were included in the final core design. Incore detector noise data were obtained during Cycle 7 at approximately 281 and 430 effective full power days (EFPD). In addition to the incore test detector signals, noise signals from the upper and lower chambers of the four excore power range detectors were recorded to aid in the analysis. The comparison of RMS signal levels for each incore detector and the phase relationships between detector locations within two test assemblies identified the first four fuel assembly bending modes associated with fixed end conditions

  11. Enhancement of heat transfer in HPLWR fuel assemblies

    International Nuclear Information System (INIS)

    Bastron, A.; Hofmeister, J.; Meyer, L.; Schulenberg, T.

    2005-01-01

    A study on different methods for enhancement of heat transfer in fuel assemblies for a High Performance Light Water Reactor has been performed to indicate the potential for a further increase of core outlet temperature at given cladding temperatures, or for reduction of peak cladding temperatures at the envisaged core outlet temperature. As a result, the introduction of an artificial surface roughness or the use of a staircase type grid spacer should increase the heat transfer coefficient of the coolant at the cladding surface by more than a factor of two, which will reduce the peak cladding temperature by at least 50 degC. The paper provides further details for realization of these measures. (author)

  12. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  13. ADELA - user interface for fuel charge design

    International Nuclear Information System (INIS)

    Havluj, Frantisek

    2010-01-01

    ADELA is a supporting computer code - ANDREA code add-on - for fuel batch designing and optimization. It facilitates fuel batch planning, evaluation and archival by using graphical user interface. ADELA simplifies and automates the design process and is closely linked to the QUADRIGA system for data library creation. (author)

  14. Fuel Element Mechanical Design for CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Estevez, Esteban; Markiewicz, Mario; Gerding, Jose

    2000-01-01

    The Fuel Element mechanical design and spider-control reactivity and security rods assembly for the CAREM-25 reactor is introduced. The CAREM-25 Fuel Element has a hexagonal cross section with 127 positions, in a triangular arrangement.There are 108 positions for the fuel rods while the guide tubes and instrumentation tube occupy the 19 remaining positions.From the structural point of view, the fuel element is being composed by a framework formed by the guides and instrumentation tubes, 4 spacer grids and the upper and lower coupling pieces.The spider is a plane piece, with a central body and six radial branches in T form, which has holes where the absorber rods are fitted.The central body ends in a joint in the upper side, which allows connect the assembly whit the reactor control mechanisms.The absorber rods are made of a neutron absorber material (Ag-In-Cd) hermetically closed in a stainless steel cladding. In this work are determined, in addition to the basic design, the operational conditions, the functional requirements to be satisfied and in agreement with those, the adopted criteria and limits to avoid systematics failure during normal operation conditions. The proposed program for the verification and evaluation of design is detailed.To consolidate the design, a prototype was manufactures, based on drawings and specifications needed for its construction

  15. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    International Nuclear Information System (INIS)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D.; Choi, B.I.; Lee, H.Y.; Song, M.J.

    2004-01-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 □ under the normal condition. The off-normal condition has an environmental temperature of 40 □. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions

  16. Fuel cells principles, design, and analysis

    CERN Document Server

    Revankar, Shripad T

    2014-01-01

    ""This book covers all essential themes of fuel cells ranging from fundamentals to applications. It includes key advanced topics important for understanding correctly the underlying multi-science phenomena of fuel cell processes. The book does not only cope with traditional fuel cells but also discusses the future concepts of fuel cells. The book is rich on examples and solutions important for applying the theory into practical use.""-Peter Lund, Aalto University, Helsinki""A good introduction to the range of disciplines needed to design, build and test fuel cells.""-Nigel Brandon, Imperial Co

  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. Description of fuel element brush assembly's fabrication for 105-K west

    International Nuclear Information System (INIS)

    Maassen, D.P.

    1997-01-01

    This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions

  19. Single rod leak detection and repair of leaking or damaged fuel assemblies

    International Nuclear Information System (INIS)

    Beuneche, D.

    1986-01-01

    In some circumstances, it is necessary to perform rework operations on some fuel assemblies in order to make them reusable in reactors, movable, transportable or consistent with fuel reprocessor specifications, depending on the plant utility policy. These rework operations are of two types: - Those which consist in restoring the leak tightness of the fuel assemblies. They are made after a series of tests allowing the localization of the failed fuel rods: at first, overall leak detection is provided by monitoring primary coolant activity during reactor operation; then, during refuelling, leaking assemblies are identified by subjecting each of the assemblies scheduled for reloading to a sipping test; finally individual leaking fuel rods are singled out before the defective assemblies can be repaired, i.e. failed rods can be replaced. - Those which involve replacement of part of or the whole assembly structure (combined or not with replacement of failed fuel rods). In order to meet these two needs for rework operations, FRAGEMA has developed a full range of test and tooling systems for detecting single leaking rods in irradiated fuel assemblies and for restoring fuel assemblies to be used in PWR nuclear power plants. As an illustration of the means available, two of these systems are described

  20. Combined fuel assembly and thimble plug gripper for a nuclear reactor

    International Nuclear Information System (INIS)

    1977-01-01

    This invention relates to an apparatus for loading and unloading a fuel assembly into and from the core of a nuclear reactor and for removing and inserting control rod guide thimble plugs from and into the fuel assembly during a reactor refueling operation in substantially less time than that presently required and in a more reliable, safe and efficient manner. (UK)

  1. Molecular Assemblies of Finite Shapes: Design and Self-Assembly ...

    Indian Academy of Sciences (India)

    Assembly via Coordination · Slide 2 · Slide 3 · Slide 4 · Slide 5 · Slide 6 · Slide 7 · Slide 8 · Slide 9 · Slide 10 · Slide 11 · Slide 12 · Slide 13 · Slide 14 · Slide 15 · Slide 16 · Slide 17 · Slide 18 · Slide 19 · Slide 20 · Slide 21 · Molecular Box · Slide 23.

  2. Combined fuel assembly and thimble plug gripper for a nuclear reactor

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Satterlee, A.E.

    1978-01-01

    A combined fuel assembly and thimble plug gripper for raising and lowering a fuel assembly into a nuclear reactor core, and for lifting and lowering a thimble plug assembly into the fuel assembly is described. It includes a vertically movable mast housing a mechanism which causes pivotally mounted fingers on the bottom of the mast to be moved into and out of latching engagement with the nozzle of a fuel assembly when the mast is resting on the assembly. The mast includes a second mechanism which supports second fingers pivotally mounted thereon and actuable by a third mechanism into and out of engagement with a thimble plug assembly supporting plugs adapted to be inserted in control rod guide thimbles in the fuel assembly. The second mechanism further includes an arrangement for lowering or raising the plug assembly respectively into or out of the guide thimbles in the fuel assembly. The apparatus includes control and interlock systems which preclude operation of the mechanisms under certain prescribed conditions

  3. 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

  4. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  5. Buoyancy-driven flow excursions in fuel assemblies

    International Nuclear Information System (INIS)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-01-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations

  6. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  7. Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

    International Nuclear Information System (INIS)

    Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

    1986-06-01

    Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

  8. 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.

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

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

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

  10. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Heins, L.; Landskron, H.

    2000-01-01

    Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO 2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

  11. Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition

    Directory of Open Access Journals (Sweden)

    Waseem

    2017-01-01

    Full Text Available Fuel assembly (FA structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear power plant-1 FA skeleton at room temperature. The finite element (FE analysis has been performed using ANSYS, in order to determine the elongation of the FA skeleton as well as the location of max. stress and stresses developed in axial direction under tensile load of 9800 N or 2 g being the FA handling or lifting load [Y. Zhang et al., Fuel Assembly Design Report, SNERDI, China, 1994]. The FE model of grids, guide thimbles with dash-pots and flow holes has been developed using Shell 181. It has been observed that FA skeleton elongation values obtained through FE analysis and experiment are comparable and show linear behaviors. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Therefore, validation of the FE methodology is confirmed. The values of stresses are less than the design limit of the materials used for the grid and the guide thimble. Therefore, the structural integrity criterion of CHASNUPP-1 FA skeleton is fulfilled safely.

  12. Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition

    Science.gov (United States)

    Waseem; Siddiqui, Ashfaq Ahmad; Murtaza, Ghulam; Maqbool, Abu Baker

    2017-12-01

    Fuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear power plant-1 FA skeleton at room temperature. The finite element (FE) analysis has been performed using ANSYS, in order to determine the elongation of the FA skeleton as well as the location of max. stress and stresses developed in axial direction under tensile load of 9800 N or 2 g being the FA handling or lifting load [Y. Zhang et al., Fuel Assembly Design Report, SNERDI, China, 1994]. The FE model of grids, guide thimbles with dash-pots and flow holes has been developed using Shell 181. It has been observed that FA skeleton elongation values obtained through FE analysis and experiment are comparable and show linear behaviors. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Therefore, validation of the FE methodology is confirmed. The values of stresses are less than the design limit of the materials used for the grid and the guide thimble. Therefore, the structural integrity criterion of CHASNUPP-1 FA skeleton is fulfilled safely.

  13. Supports with apertures for fuel assemblies in a nuclear reactor

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    This invention concerns couplings for nuclear reactor fuel rod bundle and, inter alia, for holders fitted with apertures which uncover the ends of the fuel rods for inspection, and a similar operation. The difficulties arising from the use of slid cast holders of control rods and burnable poison rods, which have marked previous designs, are overcome to a large extent through the use of the invention. Under this invention a network of interpenetrating plates is provided to hold a guide tube arrangement for the control rods or the burnable poison rods. This network forms an open cellular structure that uncovers the ends of the fuel rods, for easy routine inspection. This structure is less costly and yet has the strength of the previous solid cast holders. It avoids problems of quality guaranty encountered in the fabrication of cast metal products having a complicated shape. It allows for the re-arrangement of the partially burnt up rods inside a reactor core, without requiring any structural changes in the fuel rod bundle. It also makes it possible to overcome a want of thermal efficiency that was a hindrance in the past [fr

  14. Measurement of the gamma dose rate distribution in a spent fuel assembly with a thermoluminescent detector

    International Nuclear Information System (INIS)

    Ohno, A.; Matsuura, S.

    1980-01-01

    Thermoluminescent detectors (TLDs) were applied to measurements of gamma dose rates and their distributions in spent fuel assemblies irradiated in a boiling water reactor, the Japan Power Demonstration Reactor (JPDR). Materials used for the TLD were Mg 2 SiO 4 :Tb and LiF. Feasibility of the technique was confirmed by an on-site measurement in a spent fuel pond of the JPDR. The highest gamma dose rate calibrated to a 60 Co gamma field was 4 x 10 3 R/h in the measured positions in a spent fuel assembly, with an average burnup of 5639 MWd/MTM and a cooling time of 8 yr. Reliability of the measurement by means of a TLD was tested via duplicate measurements on a fuel assembly. In the fields of nuclear safeguards and fuel management, application of TLDs will be effective in obtaining inside information of a spent fuel assembly nondestructively

  15. Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies

    Science.gov (United States)

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

    2018-02-01

    The Digital Cherenkov Viewing Device (DCVD) is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the recording of Cherenkov light produced by the assemblies. One type of verification involves comparing the measured light intensity from an assembly with a predicted intensity, based on assembly declarations. Crucial for such analyses is the performance of the prediction model used, and recently new modelling methods have been introduced to allow for enhanced prediction capabilities by taking the irradiation history into account, and by including the cross-talk radiation from neighbouring assemblies in the predictions. In this work, the performance of three models for Cherenkov-light intensity prediction is evaluated by applying them to a set of short-cooled PWR 17x17 assemblies for which experimental DCVD measurements and operator-declared irradiation data was available; (1) a two-parameter model, based on total burnup and cooling time, previously used by the safeguards inspectors, (2) a newly introduced gamma-spectrum-based model, which incorporates cycle-wise burnup histories, and (3) the latter gamma-spectrum-based model with the addition to account for contributions from neighbouring assemblies. The results show that the two gamma-spectrum-based models provide significantly higher precision for the measured inventory compared to the two-parameter model, lowering the standard deviation between relative measured and predicted intensities from 15.2 % to 8.1 % respectively 7.8 %. The results show some systematic differences between assemblies of different designs (produced by different manufacturers) in spite of their similar PWR 17x17 geometries, and possible ways are discussed to address such differences, which may allow for even higher prediction capabilities. Still, it is concluded that the gamma-spectrum-based models enable confident verification of the fuel assembly inventory at the currently used

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

    International Nuclear Information System (INIS)

    Rosa, J.M.

    1986-01-01

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

  17. ASSEMBLY DESIGN OPTIMIZATION FOR GEAR PUMP HYDRAULIC UNITS

    Directory of Open Access Journals (Sweden)

    ŞCHEAUA Fanel

    2012-09-01

    Full Text Available This paper presents a model for gear pump assembly design realized in Solid Edge V20. The aim is to highlight modelling aspects for solid part components and how to achieve an assembly from several component parts. Can be noted that computer aided design (CAD software can provide multiple options of representing various designed components, assemblies containing up to hundreds of items and part component motion simulation.

  18. Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

    1981-01-01

    The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

  19. Fuel lattice design in boiling water reactors using path relinking

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz, J. J.; Campos, Y.; Perusquia, R.; Montes, J. L.; Hernandez, J. L.

    2006-01-01

    Full text: Full text: A new system for the optimization of fuel lattice design in boiling water reactors (BWR), using the heuristic technique called Path Relinking was developed. The system starts with an initial uranium enrichment and gadolinium percent proposal. With this information, the system generates a seed fuel lattice, which it is used to perform an iterative process until an optimized fuel lattice design is achieved. The iterative process includes two steps. In the first one, we constructed a scatter set with 96 fuel lattices, each fuel lattice we called an element. Starting from this set, we build a reference set with 10 elements, which are the best elements according to the objective function. After, from remaining 86 elements, we build the 10 elements with the maximum distance with respect to reference set. During the iterative process, elements from both sets are used to generate a new element to update the reference set. In the second step, in order to improve the solution achieved up to this moment, two elements from the reference set for constructing new paths beyond to the neighbourhood space, are used. If the new element does not improve the solution, we continue working with the same reference set in the next iteration. The objective function includes both the power peaking factor and the effective multiplication factor at the beginning of the life of the fuel lattice. The principal idea is to minimize the power peaking factor and to keep the effective multiplication factor in a proposal interval. The fuel lattice designed corresponds to the bottom of the fuel assembly. Only, if fuel lattice fulfils the requirements, then it is evaluated at several burnup points. In order to calculate the parameters involved in the objective function the 2D Helios-1.5 code was used. The system was developed in an Alpha Workstation

  20. A Study on the Development of Simplified Fuel Assembly SSE/LOCA Analysis Model using Optimization Technique

    International Nuclear Information System (INIS)

    Lee, Kyou Seok; Jeon, Sang Youn; Kim, Hyeong Koo

    2009-01-01

    Under the Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) events, the fuel assembly deflection and impact force between fuel assemblies are obtained by the dynamic transient analysis for the reactor core model. The impact behavior between fuel assemblies shows non-linear characteristics, because fuel assembly shows non-linearly dynamic characteristics and its geometry is complicated. Furthermore, since a reactor core consists of a large number of fuel assemblies, the dynamic behavior of the core under the postulated events is very difficult to analyze. Therefore, it is necessary that fuel assembly model be simplified considering dynamic non-linear characteristics in core analysis. In this study, a simplified fuel assembly finite element model for 17 Type RFA has been developed using optimization technique. To obtain the simplified model, the optimization algorithm of ANSYS was used, and the model was verified by comparison with fuel assembly mechanical test results

  1. A Study on the Development of Simplified Fuel Assembly SSE/LOCA Analysis Model using Optimization Technique

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyou Seok; Jeon, Sang Youn; Kim, Hyeong Koo [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2009-05-15

    Under the Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) events, the fuel assembly deflection and impact force between fuel assemblies are obtained by the dynamic transient analysis for the reactor core model. The impact behavior between fuel assemblies shows non-linear characteristics, because fuel assembly shows non-linearly dynamic characteristics and its geometry is complicated. Furthermore, since a reactor core consists of a large number of fuel assemblies, the dynamic behavior of the core under the postulated events is very difficult to analyze. Therefore, it is necessary that fuel assembly model be simplified considering dynamic non-linear characteristics in core analysis. In this study, a simplified fuel assembly finite element model for 17 Type RFA has been developed using optimization technique. To obtain the simplified model, the optimization algorithm of ANSYS was used, and the model was verified by comparison with fuel assembly mechanical test results.

  2. Design of A Center Deviation Adaptive Bearing Pressure Assembly Machine

    OpenAIRE

    Zhibin Chang; Lei Zhang; Guangguo Zhang

    2012-01-01

    A car transmission in order to solve the production of eight gear and nine file two models of the transmission of tapered roller bearing outer ring manual assembly accuracy, low noise, the efficiency is low, the labor intensity and difficult to realize automatic assembly line requirements so assembly problem, put forward the center deviation adaptive bearing pressure assembly machine development. Through the research, design, test and making, eventually developed used in automobile gearbox be...

  3. Nuclear Analysis for Application of Boron Burnable Absorber in the HANARO Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Choi, Wonwoo; Chae, Hee-Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seo, Keun Ho [Seoul National University, Seoul (Korea, Republic of)

    2016-10-15

    Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, and the first full core analysis was performed last year. In the analysis, cadmium in the form of CdO was considered as the most promising burnable absorber for the current HANARO element. The U-Mo fuel with CdO was successfully irradiated at the HANARO core, but it was found that the thermodynamic stability of CdO is questionable under the higher temperature condition than the current manufacturing environment. Prior to application of CdO, further studies are required. Traditionally, boron has been used as burnable absorber at the high performance research reactors such as ATR, FRM-II, etc. The power density of HANARO is lower than those reactors, and the residual reactivity effect by boron is not negligible in the core. Basic nuclear analysis for application of boron burnable absorber in the HANARO fuel assembly was performed for getting a better fuel element. The residual reactivity effect can be overcome with replacing the reduced fuel rods into the standard rods. This new fuel design provides lower reactivity swing and lower power peaking. To minimize the residual reactivity effect, a concept of heterogeneous burnable poison within the rod is introduced. The heterogeneous cases give us better results and there is an optimum boundary for the burnable absorber region. This neutronics study was limited to the boron burnable absorber in the current silicide HANARO fuel, other studies including manufacturing study are desirable for application of burnable absorber.

  4. 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

  5. Gripper assembly for inserting and removing burnable absorber rods and thimble plugs in a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Meuschke, R.E.

    1988-01-01

    This patent describes combination with elongated members, each elongated member having a releasable latching structure disposed at one end thereof, each latching structure having at least one latching member adapted to be movable between a latched position in which the latching member is able to releasably engage the adapter plate of the top nozzle of a nuclear reactor fuel assembly and secure an elongated member in a stationary relationship with respect to the adapter plate, and an unlatched position in which the latching member is able to disengage from the adapter plate so that the elongated member can be removed from the fuel assembly, apparatus for inserting and removing the elongated members from the nuclear reactor fuel assembly comprising: a frame; first and second plates disposed within the frame, the second plate disposed below and spaced apart from the first plate, at least one of the first and second plates being capable of vertical movement relative to the other; means for moving the frame toward and away from the fuel assembly; means for moving the first and second plates within the frame toward and away from the adapter plate of the top nozzle of the fuel assembly; means for varying the vertical distance between the first and second plates between a first distance and a decreased second distance; means associated with the first and second plates for maintaining the first and second plates in a position whereby the vertical distance between the plates is the second distance

  6. The application of neural networks for optimization of the configuration of fuel assemblies in PWR reactors

    International Nuclear Information System (INIS)

    Sadighi, M.; Setayeshi, S.; Salehi, A.A.

    2002-01-01

    This paper presents a new method to solve the problem of finding the best configuration of fuel assemblies in a PWR (Pressurized Water Reactor) core. Finding an optimum solution requires a huge amount of calculations in classical methods. It has been shown that the application of continuous Hop field neural network accompanied by the Simulated Annealing method to this problem not only reduces the volume of the calculations, but also guarantees finding the best solution. In this study flattening of neutron flux inside the reactor core of Brusher NPP is considered as an objective function. The result shows the optimum core configuration which is in agreement with the pattern proposed by the designer

  7. Experimental study and comparison of various designs of gas flow fields to PEM fuel cells and cell stack performance

    Directory of Open Access Journals (Sweden)

    Hong eLiu

    2014-01-01

    Full Text Available In this study, a significant number of experimental tests to PEM fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells all have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still favorable, giving the best single fuel cell performance amongst all the studied flow channel designs. A novel symmetric serpentine flow field was proposed for relatively large size fuel cell application. Four fuel cell stacks each including four cells were assembled using different designs of serpentine flow channels. The output power performances of fuel cell stacks were compared and the novel symmetric serpentine flow field design is recommended for its very good performance.

  8. Interim design report: fuel particle crushing

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

    1977-11-01

    The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

  9. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  10. A versatile passive and active non-destructive device for spent fuel assemblies monitoring

    International Nuclear Information System (INIS)

    Berne, R.; Bignan, G.; Andrieu, G.; Dethan, B.

    1993-01-01

    The monitoring of spent fuel assemblies in reactor pools or in reprocessing plants with NDA methods is interesting (non-destructivity, non-intrusivity) for process control, safety-criticality and/or nuclear material management. In this context, the authors present the results of the development and design of a prototype device (physical methods used, qualification...) called PYTHON. The aim of PYTHON is to check the declared characteristic values of an irradiated assembly before taking it into a transport cask for safety criticality control. The PYTHON device consists of a detector head in two sections and a 252 Cf source if active neutron counting is to be used. Each section of the detection head consists of two detectors: one fission chamber and one ionization chamber

  11. Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

    International Nuclear Information System (INIS)

    Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

    2008-01-01

    An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

  12. Substantiation of strength of TVSA-ALPHA fuel assembly under dynamic seismic loads

    International Nuclear Information System (INIS)

    Tutnov, A.; Kiselev, A.; Kiselev, A.; Krutko, E.; Kiselev, I.; Samoilov, O.; Kaydalov, V.

    2009-01-01

    A special place in the substantiation of the safe operation of fuel assemblies is the assessment their operating capability under seismic loads, leading to short-term (several seconds or tens of seconds) the dynamic effects on the reactor core. The level of acceleration of various elements of the reactor installation can be higher than 1,5 g (g - acceleration of gravity) and depends on the height of these elements relatively the ground, which movement causes an earthquake. This dynamic load cause significant deformation of the active zones design element, in particular of the fuel assemblies (FA), which could lead to a contact (or impact) interaction between them. The report presents the results of studies of stress-strain state of FA of TVSA-ALPHA type under the influence of seismic loads of the 8th level on Richter scale using standard approach. According to a normative approach the natural frequencies and modes of FA are calculated in the preliminary stage. The obtained results are conservative from the point of view that in the real FA design the most loaded SG in the middle of the fuel assemblies are made in a combined with mixing grid variant, which are joint by a common rim. This increases the overall carrying capacity of SG as compared with the calculation SG model. It is also necessary to bear in mind that the dynamic (impact) loading the basic mechanical properties of the material may have a significant difference from static (standard) values. This refers in particular to the yield limit, the value of which can be several times higher than specified in the calculation

  13. Load Assembly Design of the FAST Machine

    International Nuclear Information System (INIS)

    Cucchiaro, A.; Albanese, R.; Ambrosino, G.; Brolatti, G.; Calabro, G.; Cocilovo, V.; Coletti, A.; Coletti, R.; Costa, P.; Frosi, P.; Crescenzi, F.; Crisanti, F.; Granucci, G.; Maddaluno, G.; Pericoli Ridolfini, V.; Pizzuto, A.; Rita, C.; Ramogida, G.

    2008-01-01

    The FAST [1][2] (Fusion Advanced Studies Torus) load assembly, which includes the Vacuum Vessel (VV) and its internal components, the magnet system and the poloidal field coils, is presented in this paper. FAST operates at a wide range [3][4] of parameters from high performance H-Mode (B T up to 8.5 T; I P up to 8 MA) as well as Advance Tokamak operation (I P =3 MA), and full non inductive current scenario (I P =2 MA). Helium gas at 30K is used for cooling the resistive copper magnets. That allows for a pulse duration up to 170 s (∼ 40 times τres) at 3MA/ 3.5T. To limit the TF magnet ripple ferromagnetic within acceptable values insert have been introduced inside the outboard area of the VV. The VV, segmented by 20 degree modules, is capable to accommodate 40 MW RF power systems. The machine has been designed to house10 MW Negative Neutral Beam injection (NNBI) systems. Tungsten (W) and Liquid Lithium (L-Li) have been chosen as the divertor plates material, and Argon and Neon as the injected impurities to mitigate the thermal loads

  14. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

  15. The J-2X Fuel Turbopump - Design, Development, and Test

    Science.gov (United States)

    Tellier, James G.; Hawkins, Lakiesha V.; Shinguchi, Brian H.; Marsh, Matthew W.

    2011-01-01

    Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

  16. Influence of “whirlwind” mixing grids on the critical power of WWER fuel assembly

    International Nuclear Information System (INIS)

    Selivanov, Yu.F.; Pomet'ko, R.S.; Volkov, S.E.

    2014-01-01

    The problem of optimizing the number and placement of lattices in different types assemblies is discussed. The effect of the amount of mixing lattices and their locations in assemblies on the conditions of occurrence of boiling crisis in the fuel assembly on its critical power (power of assembly in case of boiling crisis) is studied. Experiments were carried out with the use of freon as a coolant. It is recommended simultaneous use in the assembly of lattices of “whirlwind” type, well-intensifying heat exchange, and cell lattices of “pass” type (or lattices with deflectors) affecting on moving flow, provided the optimal location of lattices in the assembly [ru

  17. Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

  18. Experiences in transferring of AFA 3G fuel assembly fabrication

    International Nuclear Information System (INIS)

    Yang Xiaodong; Wu Zhiming; Luo Jiankang

    2002-01-01

    Implementation program is developed for the transferring of AFA 3G technology, together with the project management experts designated by Framatome Company, to facilitate the technology import under the guidance of strict program. Technical documents and quality insurance management documents are developed based on the full understanding of the information provided by Framatome to guide the fabrication of AFA 3G fuel elements. Technical requirement suggested by Framatome is adopted as much as possible, considering the practical process capability of YFP. The focus is the technology about fabrication difficulties in the AFA 3G technology, to insure the successful transfer of the AFA 3G fabrication technology

  19. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  20. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    Energy Technology Data Exchange (ETDEWEB)

    H.C. Maru; M. Farooque

    2003-03-01

    The program efforts are focused on technology and system optimization for cost reduction, commercial design development, and prototype system field trials. The program is designed to advance the carbonate fuel cell technology from full-size field test to the commercial design. FuelCell Energy, Inc. (FCE) is in the later stage of the multiyear program for development and verification of carbonate fuel cell based power plants supported by DOE/NETL with additional funding from DOD/DARPA and the FuelCell Energy team. FCE has scaled up the technology to full-size and developed DFC{reg_sign} stack and balance-of-plant (BOP) equipment technology to meet product requirements, and acquired high rate manufacturing capabilities to reduce cost. FCE has designed submegawatt (DFC300A) and megawatt (DFC1500 and DFC3000) class fuel cell products for commercialization of its DFC{reg_sign} technology. A significant progress was made during the reporting period. The reforming unit design was optimized using a three-dimensional stack simulation model. Thermal and flow uniformities of the oxidant-In flow in the stack module were improved using computational fluid dynamics based flow simulation model. The manufacturing capacity was increased. The submegawatt stack module overall cost was reduced by {approx}30% on a per kW basis. An integrated deoxidizer-prereformer design was tested successfully at submegawatt scale using fuels simulating digester gas, coal bed methane gas and peak shave (natural) gas.

  1. Burnup credit in the design of spent-fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Wilmot, E.L.

    1987-01-01

    The spent-fuel carrying capacities of previous generation shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus, a considerable weight margin is available to the designer for increasing the payload capacity. One method of achieving an increase in capacity is by reducing fuel assembly spacing. The amount of reduction in assembly spacing available is limited by criticality and fuel support structural concerns. The optimum fuel assembly achievable is then limited by requirements to control neutron multiplication and to ensure the structural integrity of fuel support components. An investigation of the feasibility of accounting for fuel burnup in the design of spent-fuel shipping casks was recently completed for the US Dept. of Energy's Office of Civilian Radioactive Waste Management. Criticality analyses have been performed, and potential impacts in terms of increased cask capacities with associated costs and safety benefits have been determined. The scope of the analysis is limited to typical burnups of full-cycle, discharged pressurized water reactor (PWR) fuel and generic shipping cask designs. A sensitivity analysis was performed to estimate the impact of cask capacity on total transportation system life cycle costs

  2. Design of fuel element for RA10

    International Nuclear Information System (INIS)

    Estevez, Esteban A.; Markiewicz, Mario; Gerding, Roberto

    2012-01-01

    The RA-10 reactor is an open pool multipurpose reactor. It is intended for radioisotopes production, fuel irradiation and use of neutron beam experiments. The nominal configuration core consists of 19 fuel elements (FE) and 6 in-core irradiation positions. With regard to the FE, although both conceptual design and manufacturing technology are similar to the already developed and qualified by CNEA (MTR fuel flat plate), the conditions imposed by the new reactor on FE's are more demanding that previous supplies. Here it should be mentioned the magnitude of the hydrodynamic forces acting on the FE caused by coolant flow through the core (upward) and mainly by the high coolant velocity between fuel plates (greater than 5 times than those currently in operation). Moreover, the high power density results in higher heat flux in fuel plates and greater temperature gradient. As a result of these increased demands present during irradiation, and in order to maintain a high level of reliability, it is necessary carry out some modifications in the mechanical design of the FE (with respect to the so-called ECBE design or s tandard ) . Design verification is performed through analytical and code calculations, and hydrodynamic tests on a full-scale prototype. This article describes the design of the FE for RA 10 reactor, with special emphasis on those aspects that represent innovations in the traditional design (ECBE). It also presents the functional requirements, design criteria and design limits established according to the reactor operational states (author)

  3. Strength evaluation of top nozzle holddown spring screw for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Koh, S. K.; Won, S. Y.; Ryu, C. H.; Kim, Y. J.; Lee, K. S.; Jeon, K. L.

    2002-01-01

    Holddown springs are required to maintain the nuclear fuel assembly in contact with lower core plate and permit thermal and irradiation-induced length changes. Therefore, the holddown spring screw must be designed such that it is capable of sustaining the loads imposed by the initial tensile preload and operational loads. Prior to assessing the structural integrity of the spring screw in the corrosive and irradiating environment throughout the design lifetime of the fuel assembly, the strength evaluation of screw was made in this paper using the mechanics of materials and finite element methods. Calculations based on the mechanics of materials, showed that the preloaded screw with an operating holddown force had a quite large margin of safety in strength. However, the elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Preloading on the screw applied for tightening had beneficial effects on the screw strength by reducing the stress level at the critical regions, compared to the screw without preload. Calculated spring deflection using the finite element analysis was in close agreement with the experimentally measured deflection

  4. 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.

  5. 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.

  6. 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

  7. CFD analysis of coolant channel geometries for a tightly packed fuel rods assembly at subcritical pressure

    International Nuclear Information System (INIS)

    Guo, Rui; Oka, Yoshiaki

    2015-01-01

    Highlights: • Three coolant channel geometries are proposed for high breeding LWRs. • Thermal hydraulic performance of coolant channel geometries are analyzed. • Design ranges of PWRs and BWRs with proposed geometries are designated. • One type of geometry (Geometry B) is proved to be superior to the others. - Abstract: This paper analyzes the thermal hydraulic performance of channels with different cross sectional geometries, which were adopted by tightly packed fuel rods assembly for high breeding at operating pressure of PWR and BWR. The calculations were carried out to assess the geometrical effect of coolant channels on thermal-hydraulic parameters by using a CFD code STAR-CCM+. It is found that one type of channel geometry (geometry B) is superior to others because of broad design area of power, cladding temperature and pressure drop

  8. Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

    International Nuclear Information System (INIS)

    Butkovich, T.R.

    1980-05-01

    A generic test of the geologic storage of spent-fuel assemblies is being made at Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes, ADINA and ADINAT

  9. Recent improvements in the design and manufacture of LWR fuel

    International Nuclear Information System (INIS)

    Jurcevic, M.; Pevec, D.

    1998-01-01

    Strong competition among the nuclear fuel vendors on the LWR fuel market is the main driving force for improvements in the design and manufacture of the LWR fuel. Nuclear utilities are requesting more advanced fuel, which will increase the fuel availability and lower fuel cycle cost. The fuel vendors' answer to this request has been an introduction of several improvements in the design and manufacture of their fuel. In this article the current trends in advanced fuel designs and development of improved materials, which have improved the fuel utilization and availability, are discussed in more detail. Also, the impact of these improved features on the safety analyses has been evaluated.(author)

  10. Design of the MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Johnson, J.V.; Brabazon, E.J.

    2001-01-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  11. 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

  12. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

  13. 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.

  14. Experimental study on the impulsive load of follow fuel assembly of swimming pool research reactor

    International Nuclear Information System (INIS)

    Sun Changlong; Bo Hanliang; Jiang Shengyao; Zhang Hongchao; Ma Cang; Wang Jinhua; Qin Benke

    2005-01-01

    Through the data of impulsive load experiment on this fuel assembly, this paper obtained how the impact force and impact acceleration changed with time and other parameters, and analyzed the causation of the different impact force and impact acceleration, further more, found the disciplinarian of impulsive load. Investigation showed: the fuel assembly of electromagnetic movable coil control rod drive mechanism is firm enough to suffer from the impulsive load, so it suits using in the reactor. (authors)

  15. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    Energy Technology Data Exchange (ETDEWEB)

    H.C. Maru; M. Farooque

    2005-03-01

    The program was designed to advance the carbonate fuel cell technology from full-size proof-of-concept field test to the commercial design. DOE has been funding Direct FuelCell{reg_sign} (DFC{reg_sign}) development at FuelCell Energy, Inc. (FCE, formerly Energy Research Corporation) from an early state of development for stationary power plant applications. The current program efforts were focused on technology and system development, and cost reduction, leading to commercial design development and prototype system field trials. FCE, in Danbury, CT, is a world-recognized leader for the development and commercialization of high efficiency fuel cells that can generate clean electricity at power stations, or at distributed locations near the customers such as hospitals, schools, universities, hotels and other commercial and industrial applications. FCE has designed three different fuel cell power plant models (DFC300A, DFC1500 and DFC3000). FCE's power plants are based on its patented DFC{reg_sign} technology, where a hydrocarbon fuel is directly fed to the fuel cell and hydrogen is generated internally. These power plants offer significant advantages compared to the existing power generation technologies--higher fuel efficiency, significantly lower emissions, quieter operation, flexible siting and permitting requirements, scalability and potentially lower operating costs. Also, the exhaust heat by-product can be used for cogeneration applications such as high-pressure steam, district heating and air conditioning. Several sub-MW power plants based on the DFC design are currently operating in Europe, Japan and the US. Several one-megawatt power plant design was verified by operation on natural gas at FCE. This plant is currently installed at a customer site in King County, WA under another US government program and is currently in operation. Because hydrogen is generated directly within the fuel cell module from readily available fuels such as natural gas and

  16. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  17. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  18. Sodium flow distribution in test fuel assembly P-23B

    International Nuclear Information System (INIS)

    Taylor, J.P.S.

    1978-08-01

    Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

  19. HTGR fuel particle crusher: Mark 2 design

    International Nuclear Information System (INIS)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power

  20. HTGR fuel particle crusher: Mark 2 design

    Energy Technology Data Exchange (ETDEWEB)

    Baer, J.W.

    1979-06-01

    The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power.

  1. Programming biological operating systems: genome design, assembly and activation.

    Science.gov (United States)

    Gibson, Daniel G

    2014-05-01

    The DNA technologies developed over the past 20 years for reading and writing the genetic code converged when the first synthetic cell was created 4 years ago. An outcome of this work has been an extraordinary set of tools for synthesizing, assembling, engineering and transplanting whole bacterial genomes. Technical progress, options and applications for bacterial genome design, assembly and activation are discussed.

  2. A spectrum correction method for fuel assembly rehomogenization

    International Nuclear Information System (INIS)

    Lee, Kyung Taek; Cho, Nam Zin

    2004-01-01

    To overcome the limitation of existing homogenization methods based on the single assembly calculation with zero current boundary condition, we propose a new rehomogenization method, named spectrum correction method (SCM), consisting of the multigroup energy spectrum approximation by spectrum correction and the condensed two-group heterogeneous single assembly calculations with non-zero current boundary condition. In SCM, the spectrum shifting phenomena caused by current across assembly interfaces are considered by the spectrum correction at group condensation stage at first. Then, heterogeneous single assembly calculations with two-group cross sections condensed by using corrected multigroup energy spectrum are performed to obtain rehomogenized nodal diffusion parameters, i.e., assembly-wise homogenized cross sections and discontinuity factors. To evaluate the performance of SCM, it was applied to the analytic function expansion nodal (AFEN) method and several test problems were solved. The results show that SCM can reduce the errors significantly both in multiplication factors and assembly averaged power distributions

  3. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    International Nuclear Information System (INIS)

    Kim, Young Gyun; Kim, Young Il

    2006-12-01

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

  4. Direct fuel cell product design improvement

    Energy Technology Data Exchange (ETDEWEB)

    Maru, H.C.; Farooque, M. [Energy Research Corp., Danbury, CT (United States)

    1996-12-31

    Significant milestones have been attained towards the technology development field testing and commercialization of direct fuel cell power plant since the 1994 Fuel Cell Seminar. Under a 5-year cooperative agreement with the Department of Energy signed in December 1994, Energy Research Corporation (ERC) has been developing the design for a MW-scale direct fuel cell power plant with input from previous technology efforts and the Santa Clara Demonstration Project. The effort encompasses product definition in consultation with the Fuel Cell Commercialization Group, potential customers, as well as extensive system design and packaging. Manufacturing process improvements, test facility construction, cell component scale up, performance and endurance improvements, stack engineering, and critical balance-of-plant development are also addressed. Major emphasis of this product design improvement project is on increased efficiency, compactness and cost reduction to establish a competitive place in the market. A 2.85 MW power plant with an efficiency of 58% and a footprint of 420 m{sup 2} has been designed. Component and subsystem testing is being conducted at various levels. Planning and preparation for verification of a full size prototype unit are in progress. This paper presents the results obtained since the last fuel cell seminar.

  5. Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

  6. Methodologies to determine the Pu content of spent fuel assemblies for input nuclear material accountancy of pyroporcessing

    International Nuclear Information System (INIS)

    Lee, Taehoon; Shin, Heesung; Kim, Youngsoo; Kim, Hodong; Kwon, Taeje

    2011-01-01

    This study shows two different non-destructive approaches to determine the Pu mass of spent fuel assemblies, and the analysis results on the errors in their Pu mass. For both methods, the Cm mass of the assembly is obtained based on the neutron measurement results. The Cm ratio of the assembly is determined from the Cm mass and the Pu mass obtained by using either of the two methods. In a comparison of two methods, the second method is simpler than the first and may not need a homogeneously-mixed sample of the spent fuel assembly. On the other hand, the second approach shows larger error in the estimated Pu mass than the first one for many different spent fuel cases of various burnup, initial enrichment, and cooling times. A member state support program for the development of the IAEA safeguards approach for an engineering-scale pyroprocessing facility, which is designated as the Reference Engineering-scale Pyroprocessing Facility(REPF), has been carried out by Korea Atomic Energy Research Institute since 2008. The nuclear material accountancy of the REPF is based on the 'Cm balance' technique. The Pu content of processing materials of pyroprocessing can be determined by measuring the Cm mass of the materials and multiplying it by the Cm ratio. The spent fuel assembly is de-cladded, and the irradiated UO 2 material of the assembly is homogeneously mixed in the homogenization process in order to obtain a representative sample of the spent fuel assembly for determining the mass of Pu, U and Cm elements, as well as the Cm ratio of the campaign. The shipper-receiver difference between the nuclear power plant and HPC of REPF is determined at this point. We found that the error for the Pu mass and Cm ratio determined from the homogenized uranium oxide powder is the most critical for the determination of the material unaccounted for throughout the whole processes. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using non

  7. Mechanical characterization of pressurized water reactor fuel assemblies using laser techniques

    Science.gov (United States)

    Fardeau, Pierrette; Mattei, Alain; Vallory, Joelle

    1994-09-01

    The laboratory's test facilities are used to characterize the mechanical behavior of fuel assemblies of Pressurized Water Reactors (PWR). These structures are submitted to the excitation of the cooling fluid whose transverse and axial velocity fields are also measured. Fuel assemblies are made of long fuel rods (approximately 4 m) containing depleted uranium pellets. The rods are maintained together using structural grids located at different elevations. The LHC (Laboratoire d'Hydraulique de Coeur) is equipped with special test sections along with hydraulic loops to test these full scale mockups. In order to make measurement on the tested structure with optical devices, the test sections are equipped with a plexiglas front panel or viewports. The use of non contacting methods of measurement such as laser techniques are imposed by: (1) the inaccessibility of the structure studied due to the containment of the assembly in the test section, (2) the fact that the fluid (source of excitation) flowing around the rods should no be disturbed. A laser vibrometer equipped with optic fibers is used to make velocity measurements on the assemblies at different levels. In order to evaluate the level of vibration of the structure tested (fuel assembly or fuel rod), a differential laser vibrometer is used to measure the relative displacement of the assembly with respect to the test section or the relative displacement of the fuel rod with respect to the structural grids of the assembly. The differential technique is important because the higher structural modes of the assembly are mixed up in terms of frequency with the lower structural modes of vibration of the fuel rods.

  8. 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.

  9. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Czech Academy of Sciences Publication Activity Database

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, Karel

    2011-01-01

    Roč. 652, č. 1 (2011), s. 90-93 ISSN 0168-9002 Institutional research plan: CEZ:AV0Z10480505 Keywords : fuel assembly * spent fuel * track detector Subject RIV: JF - Nuclear Energetics Impact factor: 1.207, year: 2011

  10. 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)

  11. Qualification of the B and W Mark B fuel assembly for high burnup. First semi-annual progress report, July-December 1978

    International Nuclear Information System (INIS)

    Coleman, T.A.; Coppola, E.J.; Doss, P.L.; Uotinen, V.O.; Davis, H.H.

    1979-08-01

    Five Babcock and Wilcox standard Mark B (15 x 15) fuel assemblies are being irradiated in Duke Power Company's Oconee Unit 1 reactor under a research and development program sponsored by the U.S. Department of Energy. Valuable experimental data on fuel performance characteristics at burnups of > 40,000 MWd/mtU will be obtained from these assemblies. This information, at a duty approximately 20% greater than that achieved by typical discharged assemblies, will be used to qualify standard Mark B fuel assemblies for extended burnups. Efforts during this period included fuel cycle design and reload licensing of Oconee 1 for cycle 5, in which the assemblies are being irradiated, and nondestructive examination of the assemblies during the refueling outage between cycles 4 and 5. The Oconee 1 cycle 5 startup tests proceeded in a routine manner, and the reactor has operated with a 92% capacity factor since completion of power escalation testing on November 10, 1978. Irradiation of the fuel assemblies is currently in progress

  12. NDA measurements on spent fuel assemblies at Tihange 1 by means of the ION 1/FORK

    International Nuclear Information System (INIS)

    Carchon, R.; Smaers, G.; Verrecchia, G.P.D.; Arlt, R.; Stoyanova, I.; Satinet, J.

    1986-06-01

    This report describes field tests performed at Tihange 1 Nuclear Power Station on PWR spent fuel by means of the ION 1-FORK detector. Two detector systems and three electronics systems were used to investigate the same fuel assemblies with various burn-ups and cooling times. The purpose of the exercise was to test the performance of the instrument for as well inspection purposes as for fuel management. The results are presented and discussed. (Author)

  13. CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

    2010-01-01

    The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

  14. 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

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

    International Nuclear Information System (INIS)

    Waata, C.L.

    2006-07-01

    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

  16. Proton exchange fuel cell : the design, construction and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Heinzen, M.R.; Simoes, G.C.; Da Silva, L. [Univ. do Vale do Itajai, Sao Jose, SC (Brazil). Lab. de Pesquisa em Energia; Fiori, M.A.; Paula, M.M.S. [Univ. do Extremo Sul Catarinense, Santa Catarina (Brazil). Lab. de Sintese de Complexos Multifuncionais; Benavides, R. [Centro de Investigacion en Quimica Aplicada, Coahuila (Mexico)

    2010-07-15

    Polymer electrolyte membrane fuel cells (PEMFC) convert the chemical energy stored in the fuel directly into electrical energy without intermediate steps. The PEMFC operates at a relatively low operating temperature making it a good choice for mobile applications, but a high power density is needed in order to decrease the total weight of the vehicles. This paper presented a simple methodology to construct a PEMFC-type fuel cell, with particular reference to the gaseous diffuser, cell structure, the fixing plate, mounting bracket, gas distribution plates, and the membrane electrode assembly (MEA). The geometric design and meshing of the PEMFC were also described. The electrode was made using graphite with flow-field geometry. The PEMFC was tested for 100 hour of continuous work, during which time the current and voltage produced were monitored in order to evaluate the performance of the PEMFC. The materials used in the preparation of the fuel cell proved to be suitable. There was no loss of efficiency during the tests. The most relevant aspects affecting the PEMFC design were examined in an effort to optimize the performance of the cell. 13 refs., 6 figs.

  17. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    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)

  18. Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

    International Nuclear Information System (INIS)

    Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen

    2013-01-01

    The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)

  19. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  20. 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)

  1. 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.)

  2. Apparatus for removing and/or positioning fuel assemblies of a nuclear reactor

    International Nuclear Information System (INIS)

    Vuckovich, M.; Burkett, J.P.; Sallustio, J.

    1983-01-01

    Apparatus for positioning fuel assemblies of a nuclear reactor includes a control for a crane comprising a strain gauge connected to the crane line which raises and lowers the load. The signal from the strain gauge is compared with setpoints; which if the strain gauge signal exceeds a high-level setpoint, indicating that the movement of a fuel assembly is obstructed, the line drive is disabled. The line drive is also disabled if the strain gauge signal is less than a low-level setpoint, indicating that a fuel being deposited contacts the bottom of its slot or an obstruction. To preclude lateral movement of the fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge signal exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than a slack-line setpoint. (author)

  3. On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Mikhin, V.I.; Zhukov, A.V.

    1985-01-01

    One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements

  4. LH2 fuel tank design for SSTO

    Science.gov (United States)

    Wright, Geoff

    1994-01-01

    This report will discuss the design of a liquid hydrogen fuel tank constructed from composite materials. The focus of this report is to recommend a design for a fuel tank which will be able to withstand all static and dynamic forces during manned flight. Areas of study for the design include material selection, material structural analysis, heat transfer, thermal expansion, and liquid hydrogen diffusion. A structural analysis FORTRAN program was developed for analyzing the buckling and yield characteristics of the tank. A thermal analysis Excel spreadsheet was created to determine a specific material thickness which will minimize heat transfer through the wall of the tank. The total mass of the tank was determined by the combination of both structural and thermal analyses. The report concludes with the recommendation of a layered material tank construction. The designed system will include exterior insulation, combination of metal and organize composite matrices and honeycomb.

  5. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes the general safety requirements applicable to the design of both wet and dry spent fuel storage facilities; Section 3 deals with the design requirements specific to either wet or dry storage. Recommendations for the auxiliary systems of any storage facility are contained in Section 4; these are necessary to ensure the safety of the system and its safe operation. Section 5 provides recommendations for establishing the quality assurance system for a storage facility. Section 6 discusses the requirements for inspection and maintenance that must be considered during the design. Finally, Section 7 provides guidance on design features to be considered to facilitate eventual decommissioning. 18 refs

  6. The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

    2006-01-01

    Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

  7. Assembly for transport and storage of radioactive nuclear fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1978-01-01

    The invention concerns the self-control of coolant deficiencies on the transport of spent fuel elements from nuclear reactors. It guarantees that drying out of the fuel elements is prevented in case of a change of volume of the fluid contained in storage tanks and accumulators and serving as coolant and shielding medium. (TK) [de

  8. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jacobsson Svärd, Staffan, E-mail: staffan.jacobsson_svard@physics.uu.se; Holcombe, Scott; Grape, Sophie

    2015-05-21

    A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative

  9. Evaluation of a new welding process used to joint grids thimbles properly in 16 x 16 fuel assemblies, using the finite element method

    International Nuclear Information System (INIS)

    Schettino, Carlos Frederico Mattos; Silva, Marcio Adriano Coelho da

    2011-01-01

    The present work aims to evaluate structurally the new welding process used to join the grids to the guide thimbles properly in 16 x 16 fuel assemblies. This new process is an increase of the number of welding points, 4 to 8, between grids and guide thimbles, giving more stiffness to the whole structure. A finite element model of the fuel assembly design was generated in the program ANSYS 12.1. To build this model were used elements BEAM-4 and several spring type elements. The analysis covered specific loads and displacements, simulating the boundaries conditions found during small deflection acting on the entire structure. The method used to development this analysis was the simulation of a finite element model performing a fuel assembly with four weld points on each grid cell containing the guide thimbles, and then the results of it was compare with another model, with eight weld points on each grid cell containing the guide thimbles. The behavior of the structure under the acting displacement and the related results of the analysis, mainly the stiffness, were satisfied. The results of this analysis were used to prove that the new grid to guide thimble welding process improve the dimensional stability when submitted to loads and displacements required on the fuel assembly design. The performed analysis provided INB to get more information of extreme importance, for the continuity of the development of new process of manufacturing and to improve the design of the current fuel assemblies used in reactors. (author)

  10. An additive manufacturing oriented design approach to mechanical assemblies

    Directory of Open Access Journals (Sweden)

    Germain Sossou

    2018-01-01

    Full Text Available Firstly introduced as a prototyping process, additive manufacturing (AM is being more and more considered as a fully-edged manufacturing process. The number of AM processes, along with the range of processed materials are expanding. AM has made manufacturable shapes that were too difficult (or even impossible to manufacture with conventional technologies. This has promoted a shift in engineering design, from conventional design for manufacturing and assembly to design for additive manufacturing (DFAM. Research efforts into the DFAM field have been mostly dedicated to part’s design, which is actually a requirement for a better industrial adoption. This has given rise to topologically optimized and/or latticed designs. However, since AM is also capable of manufacturing fully functional assemblies requiring a few or no assembly operations, there is a need for DFAM methodologies tackling product’s development more holistically, and which are, therefore, dedicated to assembly design. Considering all the manufacturing issues related to AM of assembly-free mechanisms and available post-processing capabilities, this paper proposes a top-down assembly design methodology for AM in a proactive manner. Such an approach, can be seen as the beginning of a shift from conventional design for assembly (DFA to a new paradigm. From a product’s concept and a selected AM technology, the approach first provides assistance in the definition of the product architecture so that both functionality and successful manufacturing (including post-processing are ensured. Particularly, build-orientation and downstream processes’ characteristics are taken into account early in the design process. Secondly, for the functional flow (energy, material, signal to be appropriately conveyed by the right amount of matter, the methodology provides guidance into how the components can be designed in a minimalism fashion leveraging the shape complexity afforded by AM. A mechanical

  11. European R D programs on innovative fuel designs

    Energy Technology Data Exchange (ETDEWEB)

    Millet, P.; Languille, A. (Commissariat a l' Energie Atomique, Saint-Paul-les-Durance (France)); Brown, C. (United Kingdom Atomic Energy Authority, Caithness (United Kingdom)); Muehling, G. (Kernforschungsanlage Juelich (Germany))

    1992-01-01

    Innovative fuel design studies being carried out in Europe are designed to improve the economy (fuel cycle cost reduction) and safety of future fast reactors. This paper reports on the main research and development (R D) programs under consideration for the following: (1) dense fuels and more particularly nitride fuels; (2) advanced fuel concepts such as axially heterogeneous, vented, and fuel moderated; and (3) advanced cladding materials, especially oxide dispersion strengthened (ODS).

  12. Computationally designed peptides for self-assembly of nanostructured lattices.

    Science.gov (United States)

    Zhang, Huixi Violet; Polzer, Frank; Haider, Michael J; Tian, Yu; Villegas, Jose A; Kiick, Kristi L; Pochan, Darrin J; Saven, Jeffery G

    2016-09-01

    Folded peptides present complex exterior surfaces specified by their amino acid sequences, and the control of these surfaces offers high-precision routes to self-assembling materials. The complexity of peptide structure and the subtlety of noncovalent interactions make the design of predetermined nanostructures difficult. Computational methods can facilitate this design and are used here to determine 29-residue peptides that form tetrahelical bundles that, in turn, serve as building blocks for lattice-forming materials. Four distinct assemblies were engineered. Peptide bundle exterior amino acids were designed in the context of three different interbundle lattices in addition to one design to produce bundles isolated in solution. Solution assembly produced three different types of lattice-forming materials that exhibited varying degrees of agreement with the chosen lattices used in the design of each sequence. Transmission electron microscopy revealed the nanostructure of the sheetlike nanomaterials. In contrast, the peptide sequence designed to form isolated, soluble, tetrameric bundles remained dispersed and did not form any higher-order assembled nanostructure. Small-angle neutron scattering confirmed the formation of soluble bundles with the designed size. In the lattice-forming nanostructures, the solution assembly process is robust with respect to variation of solution conditions (pH and temperature) and covalent modification of the computationally designed peptides. Solution conditions can be used to control micrometer-scale morphology of the assemblies. The findings illustrate that, with careful control of molecular structure and solution conditions, a single peptide motif can be versatile enough to yield a wide range of self-assembled lattice morphologies across many length scales (1 to 1000 nm).

  13. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    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

  15. 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.

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

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    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 2 and UO 2 ), typically containing 95% or more UO 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 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 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

  17. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  18. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    International Nuclear Information System (INIS)

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-01-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  19. MOLTEN CARBONATE FUEL CELL PRODUCT DESIGN IMPROVEMENT

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2000-01-01

    The FCE PDI program is designed to advance the carbonate fuel cell technology from the current full-size field test to the commercial design. The specific objectives selected to attain the overall program goal are: Define power plant requirements and specifications; Establish the design for a multifuel, low-cost, modular, market-responsive power plant; Resolve power plant manufacturing issues and define the design for the commercial-scale manufacturing facility; Define the stack and balance-of-plant (BOP) equipment packaging arrangement, and module designs; Acquire capability to support developmental testing of stacks and critical BOP equipment to prepare for commercial design; and Resolve stack and BOP equipment technology issues, and design, build and field test a modular prototype power plant to demonstrate readiness for commercial entry.

  20. Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

    International Nuclear Information System (INIS)

    Dawson, D.M.; Guerra, G.; Neider, T.; Shih, P.

    1995-12-01

    This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity or other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose

  1. Testing plutonium fuel assembly production for fast-neutron reactors

    International Nuclear Information System (INIS)

    Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

    1975-01-01

    The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

  2. A code for structural analysis of fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-08-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each tubular structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equilavent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  3. Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

    Science.gov (United States)

    Dettling, Charles J.; Terry, Peter L.

    1985-03-19

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  4. Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

    Science.gov (United States)

    Dettling, Charles J.; Terry, Peter L.

    1988-03-22

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  5. Fuel assembly leak tightness control on WWER-1000 reactor

    International Nuclear Information System (INIS)

    Ivanova, R.; Gerchev, N.; Mateev, A.

    2001-01-01

    The main index for integrity of the fuel rods cladding is the specific activity value of the primary coolant. This value determines the safe operation of the reactor. The limit for safe operation of WWER-1000 reactor is the value of the total activity of Iodine isotopes in the primary coolant 5.0x10 -3 Ci/l. The paper briefly describes the methodology for performing a fuel tightness test (sipping test) and shows the results from these tests performed during the period 1987 -1999 in units 5 and 6 at the Kozloduy NPP. An additional index related to the safe operation is defined to characterize the fuel cladding integrity Fuel Reliability Index (FRI). The FRI is defined as value of the average activity of 131 I in the primary coolant, corrected with a part of precipitated 235 U migration and fixed to the general permanent purification frequency. Two criteria (quantitative and statistic) are determined to qualify the fuel cladding integrity. The results from sipping tests show good reliability of the fuel irradiated in unit 5 and 6 at the Kozloduy NPP

  6. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    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) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features 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 Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  7. Design, simulation and experimental investigation of a novel reconfigurable assembly fixture for press brakes

    OpenAIRE

    Olabanji, Olayinka; Mpofu, Khumbulani; Battaïa, Olga

    2016-01-01

    A reconfigurable assembly fixture is a major and important component of a reconfigurable assembly system. It isrequired for the assembly of a variety of press brake models inorder to reduce the assembly time and overall production time.The stages and requirements for the design of an assembly fixtureand understanding of the assembly process for press brakemodels were used to design a reconfigurable assembly fixture.A detailed design analysis of parts of the fixture and the hydraulicsystem is ...

  8. Rational assembly of nanoparticle superlattices with designed lattice symmetries

    Energy Technology Data Exchange (ETDEWEB)

    Gang, Oleg; Lu, Fang; Tagawa, Miho

    2017-09-05

    A method for lattice design via multivalent linkers (LDML) is disclosed that introduces a rationally designed symmetry of connections between particles in order to achieve control over the morphology of their assembly. The method affords the inclusion of different programmable interactions within one linker that allow an assembly of different types of particles. The designed symmetry of connections is preferably provided utilizing DNA encoding. The linkers may include fabricated "patchy" particles, DNA scaffold constructs and Y-shaped DNA linkers, anisotropic particles, which are preferably functionalized with DNA, multimeric protein-DNA complexes, and particles with finite numbers of DNA linkers.

  9. Study on an Extraction Method for a Fuel Rod Image and a Visualization of the Color Information in a Sectional Image of a Spent Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Ji Woon; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Youn, Cheung [Chungnam National University, Daejeon (Korea, Republic of)

    2007-10-15

    Image processing methods for an extraction of a nuclear fuel rod image and visualization methods of the RGB color data were studied with a sectional image of spent fuel assembly. The fuel rod images could be extracted by using a histogram analysis, an edge detection and RGB rotor data. In these results, a size of the spent fuel assembly could be measured by using a histogram analysis method and a shape of the spent fuel rod could be observed by using an edge detection method. Finally, a various analyses were established for status of the spent fuel assembly by realized various 3D images for the color data in an image of a spent fuel assembly

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

    International Nuclear Information System (INIS)

    1990-03-01

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

  11. Assembly-level analysis of heterogeneous Th–Pu PWR fuel

    International Nuclear Information System (INIS)

    Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

    2017-01-01

    Highlights: • We directly compare homogeneous and heterogeneous Th–Pu fuel. • Examine whether there is an increase in Pu incineration in the latter. • Homogeneous fuel was able to achieve much higher Pu incineration. • In the heterogeneous case, U-233 breeding is faster (larger power fraction), thus decreasing incineration of Pu. - Abstract: This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of 233 U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, 233 U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, 233 U build-up is slower in the homogeneous case and therefore Pu burning is greater. This 233 U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to burn.

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

    International Nuclear Information System (INIS)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    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 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 2 powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be