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.

Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

International Nuclear Information System (INIS)

Hermansky, J.; Petenyi, V.; Zavodsky, M.

2009-01-01

The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

Fabrication development of full-sized components for GCFR core assemblies

International Nuclear Information System (INIS)

Lindgren, J.R.; Flynn, P.W.; Foster, L.C.

1980-05-01

This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed

Reconstitution of fuel assemblies and core components

Energy Technology Data Exchange (ETDEWEB)

Hummel, Wolfgang; Langenberger, Jan [AREVA NP GmbH (Germany)

2012-11-01

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

Mixed PWR core loadings with inert matrix Pu-fuel assemblies

International Nuclear Information System (INIS)

Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

1999-01-01

The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

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

International Nuclear Information System (INIS)

Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

2000-01-01

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

Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

International Nuclear Information System (INIS)

Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

2006-01-01

The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

International Nuclear Information System (INIS)

Hursin, Mathieu; Downar, Thomas J.; Yoon, Joo Il; Joo, Han Gyu

2016-01-01

Highlights: • Reactivity initiated accident analysis with direct whole core transient transport code. • Comparison with usual “two steps” procedure. • Effect of effective delayed neutron fraction definition on energy deposition in the fuel. • Effect of homogenized few-group cross sections generation at the assembly level on energy deposition in the fuel. • Effect of effective fuel temperature definition on energy deposition in the fuel. - Abstract: The impact of the approximations in the “two-steps” procedure used in the current generation of nodal simulators for core transient calculations is assessed by using a higher order solution obtained from a direct, whole core, transient transport calculation. A control rod ejection accident in an idealized minicore is analyzed with PARCS, which uses the two-steps procedure and DeCART which provides the higher order solution. DeCART is used as lattice code to provide the homogenized cross sections and kinetics parameters to PARCS. The approximations made by using (1) the homogenized few-group cross sections and kinetic parameters generated at the assembly level, (2) an effective delayed neutrons fraction, (3) an effective fuel temperature and (4) the few-group formulation are investigated in terms of global and local core power behavior. The results presented in the paper show that the current two-steps procedure produces sufficiently accurate transient results with respect to the direct whole core calculation solution, provided that its parameters are carefully generated using the prescriptions described in the present article.

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

Todt, W.H.; Playfoot, K.C.

1979-01-01

Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length

Preparation of paclitaxel/chitosan co-assembled core-shell nanofibers for drug-eluting stent

Energy Technology Data Exchange (ETDEWEB)

Tang, Jing; Liu, Yongjia [Instrumental Analysis Center, Shanghai Jiao Tong University, 200240 Shanghai (China); State Key Laboratory of Metal Matrix Composites, School of Chemistry and Chemical Engineering, Shanghai Jiao Tong University, 200240 Shanghai (China); Zhu, Bangshang, E-mail: bshzhu@sjtu.edu.cn [Instrumental Analysis Center, Shanghai Jiao Tong University, 200240 Shanghai (China); State Key Laboratory for Modification of Chemical Fibers and Polymer Materials, Donghua University, 201620 Shanghai (China); Su, Yue; Zhu, Xinyuan [State Key Laboratory of Metal Matrix Composites, School of Chemistry and Chemical Engineering, Shanghai Jiao Tong University, 200240 Shanghai (China)

2017-01-30

Highlights: • The core-shell nanofibers (NFs) were made by the co-assembly of paclitaxel (PTX) and chitosan(CS). • The PTX/CS NFs have high PTX loading content, slow drug release and low adherence of platelets. • The PTX/CS NFs have low cytotoxicity and good haemocompatibility. • The PTX/CS NFs which could be easily coated on stents could have potential application for drug eluting stents. - Abstract: The paclitaxel/chitosan (PTX/CS) core-shell nanofibers (NFs) are easily prepared by co-assembly of PTX and CS and used in drug-eluting stent. The mixture solution of PTX (dissolved in ethanol) and CS (dissolved in 1% acetic acid water solution) under sonication will make the formation of NFs, in which small molecule PTX co-assembles with biomacromolecular CS through non-covalent interactions. The obtained NFs are tens to hundreds nanometers in diameter and millimeter level in length. Furthermore, the structure of PTX/CS NFs was characterized by confocal laser scanning microscopy (CLSM), zeta potential, X-ray photoelectron spectroscopy (XPS) and nanoscale infra-red (nanoIR), which provided evidences demonstrated that PTX/CS NFs are core-shell structures. The ‘shell’ of CS wrapped outside of the NFs, while PTX is located in the core. Thus it resulted in high drug loading content (>40 wt.%). The well-controlled drug release, low cytotoxicity and good haemocompatibility were also found in drug carrier system of PTX/CS NFs. In addition, the hydrophilic and flexible properties of NFs make them easily coating and filming on stent to prepare drug-eluting stent (DES). Therefore, this study provides a convenient method to prepare high PTX loaded NFs, which is a promising nano-drug carrier used for DES and other biomedical applications. The possible molecular mechanism of PTX and CS co-assembly and core-shell nanofiber formation is also explored. Statement of significance: We develop a convenient and efficient approach to fabricate core-shell nanofibers (NFs) through

Fuel assembly and nuclear reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

1995-01-01

The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2001-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

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

Report of lower endplug welding, and testing and inspecting result for MONJU 1{sup th} reload core fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kajiyama, Takasi; Numata, Kazuaki; Ohtani, Seiji [Quality Assuranse Section, Technical Administration Division, Plutonium Fuel Center, Tokai Works, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan); Kobayashi, Hiromi; Watanabe, Hiroaki; Goto, Tatsuro; Takahashi, Hideki; Nagasaku, Katsuhiko [Inspection Development Campany Ltd., Tokai, Ibaraki (Japan)

2000-02-01

The procedure and result of lower endplugwelding, Test and Inspection and Shipment of the 1{sup th} reload core fuel assembly (80 Fuel Assemblies) for the fast breeder reactor MONJU are reported, which had been examined and inspected in Tamatsukuri Branch, Material Insurance Office, Quality Assurance Section, Technical Administration Division, Plutonium Fuel Center (before: Inspection Section, Plutonium Fuel Division), from June 1994 to January 1996. The number of cladding tubes welded to the endplug were totally 13,804: 7,418 for Core - Inside of 43 fuel Assemblies and 6,836 for Core-Outside of 37 fuel Assemblies. 13,794 of them, 7,414 Core-Inside and 6,379 Core-Outside, were approved by the test and sent to Plutonium Fuel Center. 10 of them weren't approved mainly because of default welding. Disapproval rating was 0.07%. (author)

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

1981-01-01

An in-core neutron detector assembly consisting of a number of longitudinally extending self-powered detectors is described. The uniform mechanical structures and materials are placed symmetrically at each active detector portion thus ensuring that local perturbation factors are uniform. (U.K.)

Use of compensation assemblies in the first core of SNR-300

International Nuclear Information System (INIS)

Billaux, M.; De Wouters, R.; Pilate, S.; Vandenberg, C.

1975-01-01

For the SNR-300 reactor, the use of thin fuel pins was limited to the first core. A direct consequence of changing from the cycle reloading scheme to a complete irradiation without refueling operation is an increase of the initial excess reactivity and plutonium investment. The new system of special assemblies conceived to compensate for the too high reactivity of the first core is described: fixed absorbers, made of B 4 C pins, and sodium diluents, consisting simply of hollow wrapper tubes [fr

Giant exchange bias and its angular dependence in Co/CoO core-shell nanowire assemblies

Energy Technology Data Exchange (ETDEWEB)

Gandha, Kinjal; Chaudhary, Rakesh P.; Mohapatra, Jeotikanta; Koymen, Ali R.; Liu, J. Ping, E-mail: pliu@uta.edu

2017-07-12

The exchange-bias field (H{sub EB}) and its angular dependence are systematically investigated in Co/CoO core-shell nanowire assemblies (∼15 nm in diameter and ∼200 nm in length) consisting of single-crystalline Co core and polycrystalline CoO shell. Giant exchange-bias field (H{sub EB}) up to 2.4 kOe is observed below a blocking temperature (T{sub EB} ∼150 K) in the aligned Co/CoO nanowire assemblies. It is also found that there is an angular dependence between the H{sub EB} and the applied magnetization direction. The H{sub EB} showed a peak at 30° between the applied field and the nanowire aligned direction, which may be attributed to the noncollinear spin orientations at the interface between the ferromagnetic core and the antiferromagnetic shell. This behavior is quantitatively supported by an analytical calculation based on Stoner–Wohlfarth model. This study underlines the importance of the competing magnetic anisotropies at the interface of Co/CoO core-shell nanowires. - Highlights: • Giant exchange bias is observed in oriented Co/CoO core-shell nanowire assemblies. • Study of angular and temperature dependence of the exchange bias effect. • Competing magnetic anisotropies at the interface of Co/CoO core-shell nanowires. • Effect of misaligned spins in FM/AFM interface on angular dependence of exchange bias. • We explain the analytical model that accounts for experimental results.

Giant exchange bias and its angular dependence in Co/CoO core-shell nanowire assemblies

International Nuclear Information System (INIS)

Gandha, Kinjal; Chaudhary, Rakesh P.; Mohapatra, Jeotikanta; Koymen, Ali R.; Liu, J. Ping

2017-01-01

The exchange-bias field (H EB ) and its angular dependence are systematically investigated in Co/CoO core-shell nanowire assemblies (∼15 nm in diameter and ∼200 nm in length) consisting of single-crystalline Co core and polycrystalline CoO shell. Giant exchange-bias field (H EB ) up to 2.4 kOe is observed below a blocking temperature (T EB ∼150 K) in the aligned Co/CoO nanowire assemblies. It is also found that there is an angular dependence between the H EB and the applied magnetization direction. The H EB showed a peak at 30° between the applied field and the nanowire aligned direction, which may be attributed to the noncollinear spin orientations at the interface between the ferromagnetic core and the antiferromagnetic shell. This behavior is quantitatively supported by an analytical calculation based on Stoner–Wohlfarth model. This study underlines the importance of the competing magnetic anisotropies at the interface of Co/CoO core-shell nanowires. - Highlights: • Giant exchange bias is observed in oriented Co/CoO core-shell nanowire assemblies. • Study of angular and temperature dependence of the exchange bias effect. • Competing magnetic anisotropies at the interface of Co/CoO core-shell nanowires. • Effect of misaligned spins in FM/AFM interface on angular dependence of exchange bias. • We explain the analytical model that accounts for experimental results.

In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

International Nuclear Information System (INIS)

Todt, W.H. Sr.; Playfoot, K.C.

1988-01-01

This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

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

Coordination of Hepatitis C Virus Assembly by Distinct Regulatory Regions in Nonstructural Protein 5A.

Directory of Open Access Journals (Sweden)

Margarita Zayas

2016-01-01

Full Text Available Hepatitis C virus (HCV nonstructural protein (NS5A is a RNA-binding protein composed of a N-terminal membrane anchor, a structured domain I (DI and two intrinsically disordered domains (DII and DIII interacting with viral and cellular proteins. While DI and DII are essential for RNA replication, DIII is required for assembly. How these processes are orchestrated by NS5A is poorly understood. In this study, we identified a highly conserved basic cluster (BC at the N-terminus of DIII that is critical for particle assembly. We generated BC mutants and compared them with mutants that are blocked at different stages of the assembly process: a NS5A serine cluster (SC mutant blocked in NS5A-core interaction and a mutant lacking the envelope glycoproteins (ΔE1E2. We found that BC mutations did not affect core-NS5A interaction, but strongly impaired core-RNA association as well as virus particle envelopment. Moreover, BC mutations impaired RNA-NS5A interaction arguing that the BC might be required for loading of core protein with viral RNA. Interestingly, RNA-core interaction was also reduced with the ΔE1E2 mutant, suggesting that nucleocapsid formation and envelopment are coupled. These findings argue for two NS5A DIII determinants regulating assembly at distinct, but closely linked steps: (i SC-dependent recruitment of replication complexes to core protein and (ii BC-dependent RNA genome delivery to core protein, triggering encapsidation that is tightly coupled to particle envelopment. These results provide a striking example how a single viral protein exerts multiple functions to coordinate the steps from RNA replication to the assembly of infectious virus particles.

In Vitro-Assembled Alphavirus Core-Like Particles Maintain a Structure Similar to That of Nucleocapsid Cores in Mature Virus

OpenAIRE

Mukhopadhyay, Suchetana; Chipman, Paul R.; Hong, Eunmee M.; Kuhn, Richard J.; Rossmann, Michael G.

2002-01-01

In vitro-assembled core-like particles produced from alphavirus capsid protein and nucleic acid were studied by cryoelectron microscopy. These particles were found to have a diameter of 420 Å with 240 copies of the capsid protein arranged in a T=4 icosahedral surface lattice, similar to the nucleocapsid core in mature virions. However, when the particles were subjected to gentle purification procedures, they were damaged, preventing generation of reliable structural information. Similarly, pu...

Feasibility of fully ceramic microencapsulated (FCM) replacement fuel assembly for OPR-1000 core fully loaded with FCM fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Lee, W.J.; Lee, K.H.; Kwon, H.; Chun, J.H.; Kim, Y.M. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Venneri, F. [Ultra Safe Nuclear Corp., Los Alamos, NM (United States)

2014-07-01

The feasibility of replacing conventional UO{sub 2} fuel assemblies (FAs) of light water reactors with accident-tolerant fully ceramic microencapsulated (FCM) FAs has been explored referencing OPR-1000, 1000MW{sub e} PWR. An optimum FCM FA design, 16x16 FCM FA with Silicon Carbide-coated Zircaloy cladding, was selected based on core-level scoping analysis for five FCM FA design candidates screened from FA-level study. For the selected FCM FA design, detailed core following analysis from initial to equilibrium cores, initially fully loaded with the FCM FAs, was carried out to quantify core physics parameters. Using these parameters, the core thermal-hydraulics and coated fuel particle performance of the FCM core was assessed, and the safety margin and accident-tolerance of the FCM core was evaluated for limiting design- and beyond design-basis-accidents. From the study, it has been demonstrated that the FCM fuel is a viable option in replacing the OPR-1000 core with enhanced safety and accident tolerance while maintaining the core neutronics, thermal-hydraulics and mechanical compatibility. (author)

Conformational Changes in the Hepatitis B Virus Core Protein Are Consistent with a Role for Allostery in Virus Assembly

Energy Technology Data Exchange (ETDEWEB)

Packianathan, Charles; Katen, Sarah P.; Dann, III, Charles E.; Zlotnick, Adam (Indiana)

2010-01-12

In infected cells, virus components must be organized at the right place and time to ensure assembly of infectious virions. From a different perspective, assembly must be prevented until all components are available. Hypothetically, this can be achieved by allosterically controlling assembly. Consistent with this hypothesis, here we show that the structure of the hepatitis B virus (HBV) core protein dimer, which can spontaneously self-assemble, is incompatible with capsid assembly. Systematic differences between core protein dimer and capsid conformations demonstrate linkage between the intradimer interface and interdimer contact surface. These structures also provide explanations for the capsid-dimer selectivity of some antibodies and the activities of assembly effectors. Solution studies suggest that the assembly-inactive state is more accurately an ensemble of conformations. Simulations show that allostery supports controlled assembly and results in capsids that are resistant to dissociation. We propose that allostery, as demonstrated in HBV, is common to most self-assembling viruses.

Statistical evaluation of the on line core monitoring effectiveness for limiting the consequences of the fuel assembly misloading event

International Nuclear Information System (INIS)

Molnar, A.; Kereszturi, A.; Temesvari, E.; Korpas, L.

2007-01-01

In WWER-440 type reactors, on line core monitoring is used for the early indication of such abnormal events like fuel assembly misloading, inadvertent misalignment of Control Assemblies, blockage of coolant channels. The paper is focusing on the assembly misloading, which can not be indicated by other measurements. A Monte Carlo method was developed and applied to evaluate the on line core monitoring effectiveness for the indication of this abnormal event during the power increase in due time, when the consequences are still acceptable. The investigations proved the satisfactory effectiveness of the online core monitoring down to 55 % power even in case when 75 % of the temperature measurements was only available (Authors)

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)

Data for ion and seed dependent fibril assembly of a spidroin core domain

Directory of Open Access Journals (Sweden)

Martin Humenik

2015-09-01

Full Text Available This data article includes size exclusion chromatography data of soluble eADF4(C16, an engineered spider silk variant based on the core domain sequence of the natural dragline silk protein ADF4 of Araneus diadematus, in combination with light scattering; the protein is monomeric before assembly. The assembled mature fibrils were visualized by transmission electron microscopy (TEM and atomic force microscopy (AFM. Sonicated fibrils were used as seeds to by-pass the nucleation lag phase in eADF4(C16 assembly. We also provide data on the sedimentation kinetics of spider silk in the presence of different NaCl concentrations revealing very slow protein aggregation in comparison to the fast assembly triggered by phosphate ions published previously [1]. Experiments in the Data article represent supporting material for our work published recently [1], which described the assembly mechanism of recombinant eADF4(C16 fibrils.

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

Impaired odour discrimination on desynchronization of odour-encoding neural assemblies

Science.gov (United States)

Stopfer, Mark; Bhagavan, Seetha; Smith, Brian H.; Laurent, Gilles

1997-11-01

Stimulus-evoked oscillatory synchronization of neural assemblies has been described in the olfactory and visual systems of several vertebrates and invertebrates. In locusts, information about odour identity is contained in the timing of action potentials in an oscillatory population response, suggesting that oscillations may reflect a common reference for messages encoded in time. Although the stimulus-evoked oscillatory phenomenon is reliable, its roles in sensation, perception, memory formation and pattern recognition remain to be demonstrated - a task requiring a behavioural paradigm. Using honeybees, we now demonstrate that odour encoding involves, as it does in locusts, the oscillatory synchronization of assemblies of projection neurons and that this synchronization is also selectively abolished by picrotoxin, an antagonist of the GABAA (γ-aminobutyric acid) receptor. By using a behavioural learning paradigm, we show that picrotoxin-induced desynchronization impairs the discrimination of molecularly similar odorants, but not that of dissimilar odorants. It appears, therefore, that oscillatory synchronization of neuronal assemblies is functionally relevant, and essential for fine sensory discrimination. This suggests that oscillatory synchronization and the kind of temporal encoding it affords provide an additional dimension by which the brain could segment spatially overlapping stimulus representations.

Research program on plasma core assembly

International Nuclear Information System (INIS)

Bernard, W.; Helmick, H.H.; Jarvis, G.A.; Plassmann, E.A.; White, R.H.

1975-05-01

An operating critical assembly having a 1.024-m-diam by 1.055-m-high cavity reflected by 0.48 m of beryllium has been constructed of residual beryllium reflector segments, control drums, drive motors, and control console from the solid core nuclear rocket development program. The critical mass for uranium distributed throughout the cavity is 16.9 kg U(93.2), which is high because of neutron undermoderation in the reflector due to porosity and poison contaminants in the beryllium and graphite reflector components. A flux trapping beryllium annulus, 0.546-m-i.d. by 0.89-m-o.d., centered in the cavity, reduced the critical mass to 6.7 kg U(93.2). This arrangement will permit operation with a UF 6 zone inside the flux trap driven by an outer uranium-graphite fuel zone. The fission density inside the flux trap is appreciably higher than in the outer fuel zone. (U.S.)

RAFT Synthesis and Self-Assembly of Free-Base Porphyrin Cored Star Polymers

Directory of Open Access Journals (Sweden)

Lin Wu

2011-01-01

Full Text Available Reversible addition fragmentation chain transfer (RAFT synthesis and self-assembly of free-base porphyrin cored star polymers are reported. The polymerization, in the presence of a free-base porphyrin cored chain transfer agent (CTA-FBP, produced porphyrin star polymers with controlled molecular weights and narrow polydispersities for a number of monomers including N, N-dimethylacrylamide (DMA and styrene (St. Well-defined amphiphilic star block copolymers, P-(PS-PDMA4 and P-(PDMA-PS4 (P: porphyrin, were also prepared and used for self-assembly studies. In methanol, a selective solvent for PDMA, spherical micelles were observed for both block copolymers as characterized by TEM. UV-vis studies suggested star-like micelles were formed from P-(PS-PDMA4, while P-(PDMA-PS4 aggregated into flower-like micelles. Spectrophotometric titrations indicated that the optical response of these two micelles to external ions was a function of micellar structures. These structure-related properties will be used for micelle studies and functional material development in the future.

Three years re-use of low burned fuel assemblies from units 1 and 2 in core loading of units 3 and 4 WWER-440 at Kozloduy NPP

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.; Spasova, V.

2006-01-01

At the end of 2002 Units 1 and 2 of Kozloduy NPP were shutdown before the end of their designed lifetime. This left a large number of assemblies yet with a significant energy resource. In 2003, 43 assemblies from Unit 1 and 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 and new 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 and 66 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. The realized core loadings of Units 3 and 4 are illustrated. The large number of low burned assemblies after the shut down of Units 1 and 2 and their utilization during 4 consecutive cycles in the cores of Units 3 and 4 both do not permit to obtain in the last 3 cycles the typical mean burn up 35-36 MWd/kg U of 3.6% discharged assemblies

Self-assembly of multiferroic core-shell particulate nanocomposites through DNA-DNA hybridization and magnetic field directed assembly of superstructures

Energy Technology Data Exchange (ETDEWEB)

Sreenivasulu, Gollapudi; Srinivasan, Gopalan, E-mail: srinivas@oakland.edu, E-mail: chavez@oakland.edu [Department of Physics, Oakland University, Rochester, MI 48309-4401 (United States); Lochbiler, Thomas A.; Panda, Manashi; Chavez, Ferman A., E-mail: srinivas@oakland.edu, E-mail: chavez@oakland.edu [Department of Chemistry, Oakland University, Rochester, MI 48309-4401 (United States)

2016-04-15

Multiferroic composites of ferromagnetic and ferroelectric phases are of importance for studies on mechanical strain mediated coupling between the magnetic and electric subsystems. This work is on DNA-assisted self-assembly of superstructures of such composites with nanometer periodicity. The synthesis involved oligomeric DNA-functionalized ferroelectric and ferromagnetic nanoparticles, 600 nm BaTiO{sub 3} (BTO) and 200 nm NiFe{sub 2}O{sub 4} (NFO), respectively. Mixing BTO and NFO particles, possessing complementary DNA sequences, resulted in the formation of ordered core-shell heteronanocomposites held together by DNA hybridization. The composites were imaged by scanning electron microscopy and scanning microwave microscopy. The presence of heteroassemblies along with core-shell architecture is clearly observed. The reversible nature of the DNA hybridization allows for restructuring the composites into mm-long linear chains and 2D-arrays in the presence of a static magnetic field and ring-like structures in a rotating-magnetic field. Strong magneto-electric (ME) coupling in as-assembled composites is evident from static magnetic field H induced polarization and low-frequency magnetoelectric voltage coefficient measurements. Upon annealing the nanocomposites at high temperatures, evidence for the formation of bulk composites with excellent cross-coupling between the electric and magnetic subsystems is obtained by H-induced polarization and low-frequency ME voltage coefficient. The ME coupling strength in the self-assembled composites is measured to be much stronger than in bulk composites with randomly distributed NFO and BTO prepared by direct mixing and sintering.

Four years Re-Use of low burned fuel assemblies from units 1 and 2 in core loadings of units 3 and 4 WWER-440 at Kozloduy NPP

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.; Spasova, V.

2006-01-01

At the end of 2002 units 1 and 2 of Kozloduy NPP were shutdown before their design life time which left a large number of assemblies yet with a significant energy resources. A decision to load these assemblies into the cores of Units 3 and 4 during the next 4 cycles has been taken. In 2003, 43 assemblies from Unit 1 cycle 23 rd and 55 assemblies from Unit 2 cycle 24 th are loaded in the cores of units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 cycle 23rd and new 55 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 cycle 23 rd and 66 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2006, the next new 54 assemblies from Unit 1 cycle 23 rd and 52 assemblies from Unit 2 cycle 24 th + 2 assemblies from Unit 3 cycle 19th are loaded in the cores of Units 3 and 4 respectively. The SPPSHB computer code system is used for development and safety assessment of the fuel loading patterns of Units 3 and 4 at Kozloduy NPP with low burned assemblies from units 1 and 2 (Authors)

Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

International Nuclear Information System (INIS)

Fox, J.N.; Wei, B.C.

1978-02-01

Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

Soft magnetic characteristics of laminated magnetic block cores assembled with a high Bs nanocrystalline alloy

Directory of Open Access Journals (Sweden)

Atsushi Yao

2018-05-01

Full Text Available This paper focuses on an evaluation of core losses in laminated magnetic block cores assembled with a high Bs nanocrystalline alloy in high magnetic flux density region. To discuss the soft magnetic properties of the high Bs block cores, the comparison with amorphous (SA1 block cores is also performed. In the high Bs block core, both low core losses and high saturation flux densities Bs are satisfied in the low frequency region. Furthermore, in the laminated block core made of the high Bs alloy, the rate of increase of iron losses as a function of the magnetic flux density remains small up to around 1.6 T, which cannot be realized in conventional laminated block cores based on amorphous alloy. The block core made of the high Bs alloy exhibits comparable core loss with that of amorphous alloy core in the high-frequency region. Thus, it is expected that this laminated high Bs block core can achieve low core losses and high saturation flux densities in the high-frequency region.

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

2009-02-23

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

Self-Assembly of Fluorescent Hybrid Core-Shell Nanoparticles and Their Application.

Science.gov (United States)

Wang, Chun; Tang, Fu; Wang, Xiaoyu; Li, Lidong

2015-06-24

In this work, a fluorescent hybrid core-shell nanoparticle was prepared by coating a functional polymer shell onto silver nanoparticles via a facile one-pot method. The biomolecule poly-L-lysine (PLL) was chosen as the polymer shell and assembled onto the silver core via the amine-reactive cross-linker, 3,3'-dithiobis(sulfosuccinimidylpropionate). The fluorescent anticancer drug, doxorubicin, was incorporated into the PLL shell through the same linkage. As the cross-linker possesses a thiol-cleavable disulfide bond, disassembly of the PLL shell was observed in the presence of glutathione, leading to controllable doxorubicin release. The silver core there provided an easily modified surface to facilitate the shell coating and ensures the efficient separation of as-prepared nanoparticles from their reaction mixture through centrifugation. Cell assays show that the prepared hybrid fluorescent nanoparticles can internalize into cells possessing excellent biocompatibility prior to the release of doxorubicin, terminating cancer cells efficiently as the doxorubicin is released at the intracellular glutathione level. Such properties are important for designing smart containers for target drug delivery and cellular imaging.

Core vocabulary in the narratives of bilingual children with and without language impairment.

Science.gov (United States)

Shivabasappa, Prarthana; Peña, Elizabeth D; Bedore, Lisa M

2017-09-22

Children with primary language impairment (PLI) demonstrate deficits in morphosyntax and vocabulary. We studied how these deficits may manifest in the core vocabulary use of bilingual children with PLI. Thirty bilingual children with and without PLI who were matched pairwise (experimental group) narrated two Spanish and two English stories in kindergarten and first grade. Core vocabulary was derived from the 30 most frequently used words in the stories of 65 and 37 typically developing (TD) first graders (normative group) for Spanish and English, respectively. The number of words each child in the experimental group produced out of the 30 identified core vocabulary words and frequency of each of the core words produced each year were analysed. Children with PLI produced fewer core vocabulary words compared to their TD peers after controlling for total words produced. This difference was more pronounced in first grade. They produced core vocabulary words less frequently in kindergarten than their TD peers. Both groups produced core vocabulary words more frequently in English than Spanish. Bilingual children with PLI demonstrate a less productive core vocabulary use compared to their TD peers in both their languages illustrating the nature of their grammatical and lexical-semantic deficits.

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

Core microstructure, morphology and chain arrangement of block copolymer self-assemblies as investigated by thermal field-flow fractionation.

Science.gov (United States)

Muza, U L; Greyling, G; Pasch, H

2018-05-28

The self-assembly of block copolymers (BCPs), as a result of solvent selectivity for one block, has recently received significant attention due to novel applications of BCPs in pharmaceuticals, biomedicine, cosmetics, electronics and nanotechnology. The correlation of BCP microstructure and the structure of the resulting self-assemblies requires advanced analytical methods. However, traditional bulk characterization techniques are limited in the quest of providing detailed information regarding molar mass (M w ), hydrodynamic size (D h ), chemical composition, and morphology for these self-assemblies. In the present study, thermal field-flow fractionation (ThFFF) is utilised to investigate the impact of core microstructure on the resultant solution properties of vesicles prepared from polystyrene-polybutadiene block copolymers (PS-b-PBd) with 1.2- and 1.4-polybutadiene blocks, respectively. As compared to investigations on the impact of the corona microstructure, the impact of core microstructure on micellar properties has largely been neglected in previous work. In N,N-dimethylacetamide (DMAc) these BCPs form vesicles having PS shells and PBd cores. D h , M w , aggregation number, and critical micelle concentration of these micelles are shown to be sensitive to the core microstructure, therefore, demonstrating the potential of microstructural differences to be used for providing tuneable pathways to specific self-assemblies. It is shown that micelles prepared from BCPs of similar PS and PBd block sizes are successfully separated by ThFFF. It is further demonstrated in this study that PS-b-PBd vesicles and PS homopolymers of identical surface chemistry (PS) and comparable D h in DMAc, can be separated by ThFFF. Copyright © 2018. Published by Elsevier B.V.

HCV Core Residues Critical for Infectivity Are Also Involved in Core-NS5A Complex Formation

Science.gov (United States)

Gawlik, Katarzyna; Baugh, James; Chatterji, Udayan; Lim, Precious J.; Bobardt, Michael D.; Gallay, Philippe A.

2014-01-01

Hepatitis C virus (HCV) infection is a major cause of liver disease. The molecular machinery of HCV assembly and particle release remains obscure. A better understanding of the assembly events might reveal new potential antiviral strategies. It was suggested that the nonstructural protein 5A (NS5A), an attractive recent drug target, participates in the production of infectious particles as a result of its interaction with the HCV core protein. However, prior to the present study, the NS5A-binding site in the viral core remained unknown. We found that the D1 domain of core contains the NS5A-binding site with the strongest interacting capacity in the basic P38-K74 cluster. We also demonstrated that the N-terminal basic residues of core at positions 50, 51, 59 and 62 were required for NS5A binding. Analysis of all substitution combinations of R50A, K51A, R59A, and R62A, in the context of the HCVcc system, showed that single, double, triple, and quadruple mutants were fully competent for viral RNA replication, but deficient in secretion of viral particles. Furthermore, we found that the extracellular and intracellular infectivity of all the mutants was abolished, suggesting a defect in the formation of infectious particles. Importantly, we showed that the interaction between the single and quadruple core mutants and NS5A was impaired in cells expressing full-length HCV genome. Interestingly, mutations of the four basic residues of core did not alter the association of core or NS5A with lipid droplets. This study showed for the first time that basic residues in the D1 domain of core that are critical for the formation of infectious extracellular and intracellular particles also play a role in core-NS5A interactions. PMID:24533158

The Expanded Core Curriculum at the Indiana School for the Blind and Visually Impaired: A Case Study

Science.gov (United States)

Ringwalt, Gail Mulholland

2013-01-01

This case study investigated how the Expanded Core Curriculum (ECC) was taught to high school students who are blind or visually impaired at the Indiana School for the Blind and Visually Impaired (ISBVI). The study focused on three students pursing different academic tracks with varying degrees of vision. The students were observed throughout…

Controlled self-assembly of multiferroic core-shell nanoparticles exhibiting strong magneto-electric effects

Energy Technology Data Exchange (ETDEWEB)

Sreenivasulu, Gollapudi; Hamilton, Sean L.; Lehto, Piper R.; Srinivasan, Gopalan, E-mail: srinivas@oakland.edu [Physics Department, Oakland University, Rochester, Michigan 48309-4401 (United States); Popov, Maksym [Physics Department, Oakland University, Rochester, Michigan 48309-4401 (United States); Radiophysics Department, Taras Shevchenko National University of Kyiv, Kyiv 01601 (Ukraine); Chavez, Ferman A. [Chemistry Department, Oakland University, Rochester, Michigan 48309-4401 (United States)

2014-02-03

Ferromagnetic-ferroelectric composites show strain mediated coupling between the magnetic and electric sub-systems due to magnetostriction and piezoelectric effects associated with the ferroic phases. We have synthesized core-shell multiferroic nano-composites by functionalizing 10–100 nm barium titanate and nickel ferrite nanoparticles with complementary coupling groups and allowing them to self-assemble in the presence of a catalyst. The core-shell structure was confirmed by electron microscopy and magnetic force microscopy. Evidence for strong strain mediated magneto-electric coupling was obtained by static magnetic field induced variations in the permittivity over 16–18 GHz and polarization and by electric field induced by low-frequency ac magnetic fields.

Monomeric RC-LH1 core complexes retard LH2 assembly and intracytoplasmic membrane formation in PufX-minus mutants of Rhodobacter sphaeroides.

Science.gov (United States)

Adams, Peter G; Mothersole, David J; Ng, Irene W; Olsen, John D; Hunter, C Neil

2011-09-01

In the model photosynthetic bacterium Rhodobacter sphaeroides domains of light-harvesting 2 (LH2) complexes surround and interconnect dimeric reaction centre-light-harvesting 1-PufX (RC-LH1-PufX) 'core' complexes, forming extensive networks for energy transfer and trapping. These complexes are housed in spherical intracytoplasmic membranes (ICMs), which are assembled in a stepwise process where biosynthesis of core complexes tends to dominate the early stages of membrane invagination. The kinetics of LH2 assembly were measured in PufX mutants that assemble monomeric core complexes, as a consequence of either a twelve-residue N-terminal truncation of PufX (PufXΔ12) or the complete removal of PufX (PufX(-)). Lower rates of LH2 assembly and retarded maturation of membrane invagination were observed for the larger and less curved ICM from the PufX(-) mutant, consistent with the proposition that local membrane curvature, initiated by arrays of bent RC-LH1-PufX dimers, creates a favourable environment for stable assembly of LH2 complexes. Transmission electron microscopy and high-resolution atomic force microscopy were used to examine ICM morphology and membrane protein organisation in these mutants. Some partitioning of core and LH2 complexes was observed in PufX(-) membranes, resulting in locally ordered clusters of monomeric RC-LH1 complexes. The distribution of core and LH2 complexes in the three types of membrane examined is consistent with previous models of membrane curvature and domain formation (Frese et al., 2008), which demonstrated that a combination of crowding and asymmetries in sizes and shapes of membrane protein complexes drives membrane organisation. 2011 Elsevier B.V. All rights reserved.

Neutronic characteristics of FLWR in the transition phase changing from high conversion core to breeder core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2009-01-01

Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation type boiling water reactor which can realize plutonium multiple recycling and breeding. For the introduction stage of FLWR, a high conversion (HC) type FLWR is proposed to keep technical continuity from current light water reactors. The HC core of FLWR has a less tight fuel lattice with lower coolant void fraction than the breeder (BR) type core. The HC type FLWR core is to be shifted to the BR core by only replacing the fuel assemblies of the same outer shape and size in the same reactor system. In the HC to BR transition phase of FLWR, there exist both types of fuel assemblies in the same core configuration. In the HC assembly, neutron spectrum is softer than in the BR assembly, and the axial fuel and blanket arrangement is different from the BR assembly. Due to these differences, there might appear a power peaking in the adjacent region between HC and BR assemblies. The power distribution in the HC + BR assemblies mixed core configuration is studied by performing assembly calculations and core calculations on a few assemblies local geometry and the whole core geometry. As a result, although a power peaking can be locally very large in the HC and BR assemblies adjacent regions, such local power peakings are shown to be effectively reduced by considering a rod-wise fuel enrichment distribution. In the whole core calculation, it seems possible to optimize the fuel assembly loading and shuffling pattern to avoid large power level mismatch between the assemblies. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties. (author)

Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

International Nuclear Information System (INIS)

Petenyi, V.; Klucarova, K.; Remis, J.

2003-01-01

The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

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

International Nuclear Information System (INIS)

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 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 rather

An experimental study of the heterogeneous LMFBR core using FCA assemblies with axial internal blanket

International Nuclear Information System (INIS)

Nakano, M.; Iijima, S.; Shirakata, K.; Hirota, J.

1980-01-01

To investigate physics properties of the heterogeneous LMFBR core and to examine the reliability of the current data and method for heterogeneous core configuration, an experimental study has been made on FCA VII-3 assemblies which have an internal blanket (IB) at midplane of the cylindrical core. Systematic experiments were carried out on the heterogeneous cores whose IBs were different in composition and thickness. A homogeneous core was also built to compare the results with those obtained on the heterogeneous cores. The sodium-void worth is not sensitive to the composition of IB. The positive void worth in the core of the 40 cm IB is lowered by about 40% compared with that in the homogeneous core. The analysis was made using the JAERI-Fast Set Version II and the diffusion code CITATION. Directional diffusion coefficients were used to take account of the axial streaming. To evaluate transport effects, the S 4 calculation was made. Comparison between the calculated and experimental results reveals the following: ksub(eff) and Pu worth in the core are not well predicted for the heterogeneous core, although they are represented satisfactorily for the homogeneous core. Reaction rates sensitive to the low-energy neutron are underestimated in the IB when they are normalized in the core. Sodium-void worths are fairly well predicted. However, the positive void worth in the heterogeneous core is underestimated, while that in the homogeneous core is overestimated. (author)

Charge neutralization as the major factor for the assembly of nucleocapsid-like particles from C-terminal truncated hepatitis C virus core protein

OpenAIRE

Theo Luiz Ferraz de Souza; Sheila Maria Barbosa de Lima; Vanessa L. de Azevedo Braga; David S. Peabody; Davis Fernandes Ferreira; M. Lucia Bianconi; Andre Marco de Oliveira Gomes; Jerson Lima Silva; Andréa Cheble de Oliveira

2016-01-01

Background Hepatitis C virus (HCV) core protein, in addition to its structural role to form the nucleocapsid assembly, plays a critical role in HCV pathogenesis by interfering in several cellular processes, including microRNA and mRNA homeostasis. The C-terminal truncated HCV core protein (C124) is intrinsically unstructured in solution and is able to interact with unspecific nucleic acids, in the micromolar range, and to assemble into nucleocapsid-like particles (NLPs) in vitro. The specific...

Classification of the NPP core and fuel assembly states by the pattern recoguition method

International Nuclear Information System (INIS)

Egorov, Yu.A.; Ivanov, E.A.; Kazakov, S.V.; Tolstykh, V.D.

1981-01-01

The patern recognition methods used for solving the problems of analysis of radiohazard states of fuel assemblies (FA) and uranium-graphite reactor core as a whole are considered. The problem under consideration is formulated as the problem of studying the deformation of signal space for the system of fuel can tightness control on the background of fuel assembly character space as characteristics, reflecting the FA living conditions in a core power, coolant flow rate, coolant steam content and pipeline length up to the detector of the system of fuel can tightness control are chosen. The analysis of deformation of the fuel can tightness control system signal space is completed by its division into two spaces: the background signal space and the valid signal space. For solving the problem the method of basic components and variational approach have been used. The conclusion is drawn that as the extent of FA failure and valid signals of by-channel system of fuel can tightness control are in one-to-one correspondence it is advantageous to solve the problem of FA state classification in the space of valid signals [ru

Charge neutralization as the major factor for the assembly of nucleocapsid-like particles from C-terminal truncated hepatitis C virus core protein.

Science.gov (United States)

de Souza, Theo Luiz Ferraz; de Lima, Sheila Maria Barbosa; Braga, Vanessa L de Azevedo; Peabody, David S; Ferreira, Davis Fernandes; Bianconi, M Lucia; Gomes, Andre Marco de Oliveira; Silva, Jerson Lima; de Oliveira, Andréa Cheble

2016-01-01

Hepatitis C virus (HCV) core protein, in addition to its structural role to form the nucleocapsid assembly, plays a critical role in HCV pathogenesis by interfering in several cellular processes, including microRNA and mRNA homeostasis. The C-terminal truncated HCV core protein (C124) is intrinsically unstructured in solution and is able to interact with unspecific nucleic acids, in the micromolar range, and to assemble into nucleocapsid-like particles (NLPs) in vitro . The specificity and propensity of C124 to the assembly and its implications on HCV pathogenesis are not well understood. Spectroscopic techniques, transmission electron microscopy and calorimetry were used to better understand the propensity of C124 to fold or to multimerize into NLPs when subjected to different conditions or in the presence of unspecific nucleic acids of equivalent size to cellular microRNAs. The structural analysis indicated that C124 has low propensity to self-folding. On the other hand, for the first time, we show that C124, in the absence of nucleic acids, multimerizes into empty NLPs when subjected to a pH close to its isoelectric point (pH ≈ 12), indicating that assembly is mainly driven by charge neutralization. Isothermal calorimetry data showed that the assembly of NLPs promoted by nucleic acids is enthalpy driven. Additionally, data obtained from fluorescence correlation spectroscopy show that C124, in nanomolar range, was able to interact and to sequester a large number of short unspecific nucleic acids into NLPs. Together, our data showed that the charge neutralization is the major factor for the nucleocapsid-like particles assembly from C-terminal truncated HCV core protein. This finding suggests that HCV core protein may physically interact with unspecific cellular polyanions, which may correspond to microRNAs and mRNAs in a host cell infected by HCV, triggering their confinement into infectious particles.

Self-assembly of star micelle into vesicle in solvents of variable quality: the star micelle retains its core-shell nanostructure in the vesicle.

Science.gov (United States)

Liu, Nijuan; He, Qun; Bu, Weifeng

2015-03-03

Intra- and intermolecular interactions of star polymers in dilute solutions are of fundamental importance for both theoretical interest and hierarchical self-assembly into functional nanostructures. Here, star micelles with a polystyrene corona and a small ionic core bearing platinum(II) complexes have been regarded as a model of star polymers to mimic their intra- and interstar interactions and self-assembled behaviors in solvents of weakening quality. In the chloroform/methanol mixture solvents, the star micelles can self-assemble to form vesicles, in which the star micelles shrink significantly and are homogeneously distributed on the vesicle surface. Unlike the morphological evolution of conventional amphiphiles from micellar to vesicular, during which the amphiphilic molecules are commonly reorganized, the star micelles still retain their core-shell nanostructures in the vesicles and the coronal chains of the star micelle between the ionic cores are fully interpenetrated.

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi; Masumi, Ryoji; Soneda, Hideo.

1994-01-01

A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison rods incorporated with burnable poisons, and water rods which can vary the height in the tube depending on the coolant flow rate flown into the assembly. The amount of entire burnable poisons of the burnable poison-containing rods in adjacent with the water rods is smaller than the amount of entire burnable poisons in the burnable poison containing rods not in adjacent with the water rods. Then the average concentration of burnable poisons in the axial upper half region is made smaller than the average concentration of the burnable poisons at the axial lower half region. Further, a burnable poison concentration at the upper half region of at least one of burnable poison-containing rods in adjacent with the water rods is made lower than the burnable poison concentration in the lower half region. Since this can fasten the combustion of the burnable poisons, a fuel assembly having good fuel economy can be attained. (I.N.)

Seismic core shroud

International Nuclear Information System (INIS)

Puri, A.; Mullooly, J.F.

1981-01-01

A core shroud is provided, comprising: a coolant boundary, following the shape of the core boundary, for channeling the coolant through the fuel assemblies; a cylindrical band positioned inside the core barrel and surrounding the coolant boundary; and support members extending from the coolant boundary to the band, for transferring load from the coolant boundary to the band. The shroud may be assembled in parts using automated welding techniques, and it may be adjusted to fit the reactor core easily

Self-Assembled Core-Shell CdTe/Poly(3-hexylthiophene) Nanoensembles as Novel Donor-Acceptor Light-Harvesting Systems.

Science.gov (United States)

Istif, Emin; Kagkoura, Antonia; Hernandez-Ferrer, Javier; Stergiou, Anastasios; Skaltsas, Theodosis; Arenal, Raul; Benito, Ana M; Maser, Wolfgang K; Tagmatarchis, Nikos

2017-12-27

The self-assembly of novel core-shell nanoensembles consisting of regioregular poly(3-hexylthiophene) nanoparticles (P3HT NPs ) of 100 nm as core and semiconducting CdTe quantum dots (CdTe QDs ) as shell with a thickness of a few tens of nanometers was accomplished by employing a reprecipitation approach. The structure, morphology, and composition of CdTe QDs /P3HT NPs nanoensembles were confirmed by high-resolution scanning transmission microscopy and dynamic light-scattering studies. Intimate interface contact between the CdTe QDs shell and the P3HT NPs core leads to the stabilization of the CdTe QDs /P3HT NPs nanoensemble as probed by the steady-state absorption spectroscopy. Effective quenching of the characteristic photoluminescence of CdTe QDs at 555 nm, accompanied by simultaneous increase in emission of P3HT NPs at 660 and 720 nm, reveals photoinduced charge-transfer processes. Probing the redox properties of films of CdTe QDs /P3HT NPs further proves the formation of a stabilized core-shell system in the solid state. Photoelectrochemical assays on CdTe QDs /P3HT NPs films show a reversible on-off photoresponse at a bias voltage of +0.8 V with a 3 times increased photocurrent compared to CdTe QDs . The improved charge separation is directly related to the unique core-shell configuration, in which the outer CdTe QDs shell forces the P3HT NPs core to effectively act as electron acceptor. The creation of novel donor-acceptor core-shell hybrid materials via self-assembly is transferable to other types of conjugated polymers and semiconducting nanoparticles. This work, therefore, opens new pathways for the design of improved optoelectronic devices.

The Replacement of 10 Non-Conserved Residues in the Core Protein of JFH-1 Hepatitis C Virus Improves Its Assembly and Secretion.

Directory of Open Access Journals (Sweden)

Loïc Etienne

Full Text Available Hepatitis C virus (HCV assembly is still poorly understood. It is thought that trafficking of the HCV core protein to the lipid droplet (LD surface is essential for its multimerization and association with newly synthesized HCV RNA to form the viral nucleocapsid. We carried out a mapping analysis of several complete HCV genomes of all genotypes, and found that the genotype 2 JFH-1 core protein contained 10 residues different from those of other genotypes. The replacement of these 10 residues of the JFH-1 strain sequence with the most conserved residues deduced from sequence alignments greatly increased virus production. Confocal microscopy of the modified JFH-1 strain in cell culture showed that the mutated JFH-1 core protein, C10M, was present mostly at the endoplasmic reticulum (ER membrane, but not at the surface of the LDs, even though its trafficking to these organelles was possible. The non-structural 5A protein of HCV was also redirected to ER membranes and colocalized with the C10M core protein. Using a Semliki forest virus vector to overproduce core protein, we demonstrated that the C10M core protein was able to form HCV-like particles, unlike the native JFH-1 core protein. Thus, the substitution of a few selected residues in the JFH-1 core protein modified the subcellular distribution and assembly properties of the protein. These findings suggest that the early steps of HCV assembly occur at the ER membrane rather than at the LD surface. The C10M-JFH-1 strain will be a valuable tool for further studies of HCV morphogenesis.

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Marmonier, Pierre; Mesnage, Bernard; Nervi, J.C.

1975-01-01

This invention refers to fuel assemblies for a liquid metal cooled fast neutron reactor. Each assembly is composed of a hollow vertical casing, of regular polygonal section, containing a bundle of clad pins filled with a fissile or fertile substance. The casing is open at its upper end and has a cylindrical foot at its lower end for positioning the assembly in a housing provided in the horizontal diagrid, on which the core assembly rests. A set of flat bars located on the external surface of the casing enables it to be correctly orientated in its housing among the other core assemblies [fr

Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

International Nuclear Information System (INIS)

Viallet, E.; Bolsee, G.; Ladouceur, B.; Goubin, T.; Rigaudeau, J.

2003-01-01

The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

Plutonium demonstration assemblies in the CNA core

International Nuclear Information System (INIS)

Romain, J.L.

1975-01-01

The SENA (Societe d'Exploitation Nucleaire des Ardennes) has decided to load two plutonium assemblies at the beginning of the 6th cycle of the CNA (Centrale Nucleaire des Ardennes). These assemblies will be loaded in September 1975 and will undergo three irradiation cycles. A general description of these two assemblies and nuclear design calculations that have been performed are presented [fr

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

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.

Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

2012-05-15

At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

Charge neutralization as the major factor for the assembly of nucleocapsid-like particles from C-terminal truncated hepatitis C virus core protein

Directory of Open Access Journals (Sweden)

Theo Luiz Ferraz de Souza

2016-11-01

Full Text Available Background Hepatitis C virus (HCV core protein, in addition to its structural role to form the nucleocapsid assembly, plays a critical role in HCV pathogenesis by interfering in several cellular processes, including microRNA and mRNA homeostasis. The C-terminal truncated HCV core protein (C124 is intrinsically unstructured in solution and is able to interact with unspecific nucleic acids, in the micromolar range, and to assemble into nucleocapsid-like particles (NLPs in vitro. The specificity and propensity of C124 to the assembly and its implications on HCV pathogenesis are not well understood. Methods Spectroscopic techniques, transmission electron microscopy and calorimetry were used to better understand the propensity of C124 to fold or to multimerize into NLPs when subjected to different conditions or in the presence of unspecific nucleic acids of equivalent size to cellular microRNAs. Results The structural analysis indicated that C124 has low propensity to self-folding. On the other hand, for the first time, we show that C124, in the absence of nucleic acids, multimerizes into empty NLPs when subjected to a pH close to its isoelectric point (pH ≈ 12, indicating that assembly is mainly driven by charge neutralization. Isothermal calorimetry data showed that the assembly of NLPs promoted by nucleic acids is enthalpy driven. Additionally, data obtained from fluorescence correlation spectroscopy show that C124, in nanomolar range, was able to interact and to sequester a large number of short unspecific nucleic acids into NLPs. Discussion Together, our data showed that the charge neutralization is the major factor for the nucleocapsid-like particles assembly from C-terminal truncated HCV core protein. This finding suggests that HCV core protein may physically interact with unspecific cellular polyanions, which may correspond to microRNAs and mRNAs in a host cell infected by HCV, triggering their confinement into infectious particles.

CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

International Nuclear Information System (INIS)

Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

2015-01-01

Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

Work-related musculoskeletal disorders (WMDs) risk assessment at core assembly production of electronic components manufacturing company

Science.gov (United States)

Yahya, N. M.; Zahid, M. N. O.

2018-03-01

This study conducted to assess the work-related musculoskeletal disorders (WMDs) among the workers at core assembly production in an electronic components manufacturing company located in Pekan, Pahang, Malaysia. The study is to identify the WMDs risk factor and risk level. A set of questionnaires survey based on modified Nordic Musculoskeletal Disorder Questionnaires have been distributed to respective workers to acquire the WMDs risk factor identification. Then, postural analysis was conducted in order to measure the respective WMDs risk level. The analysis were based on two ergonomics assessment tools; Rapid Upper Limb Assessment (RULA) and Rapid Entire Body Assessment (REBA). The study found that 30 respondents out of 36 respondents suffered from WMDs especially at shoulder, wrists and lower back. The WMDs risk have been identified from unloading process, pressing process and winding process. In term of the WMDs risk level, REBA and RULA assessment tools have indicated high risk level to unloading and pressing process. Thus, this study had established the WMDs risk factor and risk level of core assembly production in an electronic components manufacturing company at Malaysia environment.

Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

International Nuclear Information System (INIS)

Krause, F.

1986-01-01

In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

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

Integrated magnetic transformer assembly

DEFF Research Database (Denmark)

2014-01-01

The present invention relates to an integrated magnetics transformer assembly comprising a first magnetically permeable core forming a first substantially closed magnetic flux path and a second magnetically permeable core forming a second substantially closed magnetic flux path. A first input...... inductor winding is wound around a first predetermined segment of the first magnetically permeable core and a second input inductor winding is wound around a first predetermined segment of the second magnetically permeable core. The integrated magnetics transformer assembly further comprises a first output......-winding of the first output inductor winding and the first half-winding of the second output inductor winding are configured to produce aligned, i.e. in the same direction, magnetic fluxes through the first substantially closed magnetic flux path. The integrated magnetics transformer assembly is well- suited for use...

In vitro assembly into virus-like particles is an intrinsic quality of Pichia pastoris derived HCV core protein

International Nuclear Information System (INIS)

Acosta-Rivero, Nelson; Rodriguez, Armando; Musacchio, Alexis; Falcon, Viviana; Suarez, Viana M.; Martinez, Gillian; Guerra, Ivis; Paz-Lago, Dalila; Morera, Yanelys; Rosa, Maria C. de la; Morales-Grillo, Juan; Duenas-Carrera, Santiago

2004-01-01

Different variants of hepatitis C virus core protein (HCcAg) have proved to self-assemble in vitro into virus-like particles (VLPs). However, difficulties in obtaining purified mature HCcAg have limited these studies. In this study, a high degree of monomeric HCcAg purification was accomplished using chromatographic procedures under denaturing conditions. Size exclusion chromatography and sucrose density gradient centrifugation of renatured HCcAg (in the absence of structured RNA) under reducing conditions suggested that it assembled into empty capsids. The electron microscopy analysis of renatured HCcAg showed the presence of spherical VLPs with irregular shapes and an average diameter of 35 nm. Data indicated that HCcAg monomers assembled in vitro into VLPs in the absence of structured RNA, suggesting that recombinant HCcAg used in this work contains all the information necessary for the assembly process. However, they also suggest that some cellular factors might be required for the proper in vitro assembly of capsids

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)

Design Report for the core design of the first core Mark-Ia of the SNR-300

International Nuclear Information System (INIS)

Stanculescu, A.

1984-05-01

The report describes the first core Mark-Ia of the SNR-300 reactor and its different assembly types with their operational strategy. Methods, criteria and results of the neutron physical, thermal hydraulic and core mechanical design of the whole core and its assemblies are presented

TracWorks - global fuel assembly data management

International Nuclear Information System (INIS)

Cooney, B.F.

1997-01-01

The TracWorks 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 refuelling 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. (UK)

Nuclear reactor core stabilizing arrangement

International Nuclear Information System (INIS)

Jabsen, F.S.

1976-01-01

A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

DEHP exposure impairs mouse oocyte cyst breakdown and primordial follicle assembly through estrogen receptor-dependent and independent mechanisms

Energy Technology Data Exchange (ETDEWEB)

Mu, Xinyi [Laboratory of Reproductive Biology, Chongqing Medical University, Chongqing 400016 (China); Department of Histology and Embryology, College of Basic Medicine, Chongqing Medical University, Chongqing 400016 (China); Liao, Xinggui; Chen, Xuemei; Li, Yanli; Wang, Meirong; Shen, Cha; Zhang, Xue; Wang, Yingxiong; Liu, Xueqing [Laboratory of Reproductive Biology, Chongqing Medical University, Chongqing 400016 (China); He, Junlin, E-mail: hejunlin_11@aliyun.com [Laboratory of Reproductive Biology, Chongqing Medical University, Chongqing 400016 (China)

2015-11-15

Highlights: • DEHP inhibits primordial folliculogenesis in vivo and in vitro. • Estrogen receptors participate in the effect of DEHP on early ovarian development. • DEHP exposure impairs the expression of Notch2 signaling components. • DEHP exposure disrupts the proliferation of pregranulosa precursor cells. - Abstract: Estrogen plays an essential role in the development of mammalian oocytes, and recent studies suggest that it also regulates primordial follicle assembly in the neonatal ovaries. During the last decade, potential exposure of humans and animals to estrogen-like endocrine disrupting chemicals has become a growing concern. In the present study, we focused on the effect of diethylhexyl phthalate (DEHP), a widespread plasticizer with estrogen-like activity, on germ-cell cyst breakdown and primordial follicle assembly in the early ovarian development of mouse. Neonatal mice injected with DEHP displayed impaired cyst breakdown. Using ovary organ cultures, we revealed that impairment was mediated through estrogen receptors (ERs), as ICI 182,780, an efficient antagonist of ER, reversed this DEHP-mediated effect. DEHP exposure reduced the expression of ERβ, progesterone receptor (PR), and Notch2 signaling components. Finally, DEHP reduced proliferation of pregranulosa precursor cells during the process of primordial folliculogenesis. Together, our results indicate that DEHP influences oocyte cyst breakdown and primordial follicle formation through several mechanisms. Therefore, exposure to estrogen-like chemicals during fetal or neonatal development may adversely influence early ovarian development.

Analysis of PWR assembly bow

International Nuclear Information System (INIS)

Fetterman, Robert J.; Franceschini, Fausto

2008-01-01

Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

Analysis of PWR assembly bow

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J.; Franceschini, Fausto [Westinghouse Electric Company LLC, Pittsburgh, PA (United States)

2008-07-01

Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

Peripheral pin alignment system for fuel assemblies

International Nuclear Information System (INIS)

Anthony, A.J.

1981-01-01

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

Impact forces on a core shroud of an excited PWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Collard, B.; Vallory, J. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

2001-07-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment.

Impact forces on a core shroud of an excited PWR fuel assembly

International Nuclear Information System (INIS)

Collard, B.; Vallory, J.

2001-01-01

Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

Optimized core loading sequence for Ukraine WWER-1000 reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Nuclear reactor control assembly

International Nuclear Information System (INIS)

Negron, S.B.

1991-01-01

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

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

Analysis of measurements for a uranium-free core experiment at the BFS-2 critical assembly

Energy Technology Data Exchange (ETDEWEB)

Hunter, Stuart [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of Keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2D) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and Keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in Keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of Keff was 1.1%{delta}k/k higher than the measured value, Na void worth C/E values were {approx}1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes , though the effect should be investigated in any future experiments.) Several sample worth values were small compared with calculational

Core symptoms not meeting criteria for delirium are associated with cognitive and functional impairment and mood and behavior problems in older long-term care residents.

Science.gov (United States)

Cole, Martin G; McCusker, Jane; Voyer, Philippe; Monette, Johanne; Champoux, Nathalie; Ciampi, Antonio; Belzile, Eric; Vu, Minh

2014-07-01

The immediate clinical significance of Confusion Assessment Method (CAM)-defined core symptoms of delirium not meeting criteria for delirium is unclear. This study proposed to determine if such symptoms are associated with cognitive and functional impairment, mood and behavior problems and increased Burden of Care (BOC) in older long-term care (LTC) residents. The study was a secondary analysis of data collected for a prospective cohort study of delirium. Two hundred and fifty-eight LTC residents aged 65 years and older in seven LTC facilities had monthly assessments (for up to six months) of CAM - defined core symptoms of delirium (fluctuation, inattention, disorganized thinking, and altered level of consciousness) and five outcome measures: Mini-Mental State Exam, Barthel Index, Cornell Scale for Depression, Nursing Home Behavioral Problems Scale, and Burden of Care. Associations between core symptoms and the five outcome measures were analyzed using generalized estimating equations. Core symptoms of delirium not meeting criteria for delirium among residents with and without dementia were associated with cognitive and functional impairment and mood and behavior problems but not increased BOC. The associations appear to be intermediate between those of full delirium and no core symptoms and were greater for residents with than without dementia. CAM-defined core symptoms of delirium not meeting criteria for delirium appear to be associated with cognitive and functional impairment and mood and behavior problems in LTC residents with or without dementia. These findings may have implications for the prevention and management of such impairments and problems in LTC settings.

A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation

International Nuclear Information System (INIS)

Ricciardi, G.

2008-10-01

The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Masumi, Ryoji; Ishibashi, Yoko.

1995-01-01

A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison-incorporated fuel rods and a spectral shift-type water rod. As the burnable poison for the burnable poison-incorporated fuel rod, a plurality of burnable poison elements each having a different neutron absorption cross section are used. A burnable poison element such as boron having a relatively small neutron absorbing cross section is disposed more in the upper half region than the lower half region of the burnable poison-incorporated fuel rods. In addition, a burnable poison element such as gadolinium having a relatively large neutron absorbing cross section is disposed more in the lower half-region than the upper half region thereof. This can flatten the power distribution in the vertical direction of the fuel assembly and the power distribution in the horizontal direction at the final stage of the operation cycle. (I.N.)

Dynamic structural analysis for assemblies of fuel elements in the core of a PWR

International Nuclear Information System (INIS)

Silva Macedo, L.V. da.

1991-01-01

It is presented a procedure for the dynamic structural analysis of a PWR core. Impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. It is necessary a time-history response analysis. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. It is presented an algorithm of solution and also results obtained with the STYCA computer program, developed in the basis of what was proposed here. (author)

In vivo roles of BamA, BamB and BamD in the biogenesis of BamA, a core protein of the β-barrel assembly machine of Escherichia coli

Science.gov (United States)

Misra, Rajeev; Stikeleather, Ryan; Gabriele, Rebecca

2014-01-01

Assembly of the β-barrel outer membrane proteins (OMPs) is an essential cellular process in Gram negative bacteria and in the mitochondria and chloroplasts of eukaryotes—two organelles of bacterial origin. Central to this process is the conserved β-barrel OMP that belongs to the Omp85 superfamily. In Escherichia coli, BamA is the core β-barrel OMP, and together with four outer membrane lipoproteins, BamBCDE, constitute the β-barrel assembly machine (BAM). In this paper, we investigated the roles of BamD, an essential lipoprotein, and BamB in BamA biogenesis. Depletion of BamD caused impairment in BamA biogenesis and cessation of cell growth. These defects of BamD depletion were partly reversed by single amino acid substitutions mapping within the β-barrel domain of BamA. However, in the absence of BamB, the positive effects of the β-barrel substitutions on BamA biogenesis under BamD depletion conditions were nullified. By employing a BamA protein bearing one such substitution, F494L, it was demonstrated that the mutant BamA protein could not only assemble without BamD, but it could also facilitate the assembly of wild-type BamA expressed in trans. Based on these data, we propose a model in which the Bam lipoproteins, which are localized to the outer membrane by the BAM-independent Lol pathway, aid in the creation of new BAM complexes by serving as outer membrane receptors and folding factors for nascent BamA molecules. The newly assembled BAM holocomplex then catalyzes the assembly of substrate OMPs and BamA. These in vivo findings are corroborated by recently published in vitro data. PMID:24792419

In vivo roles of BamA, BamB and BamD in the biogenesis of BamA, a core protein of the β-barrel assembly machine of Escherichia coli.

Science.gov (United States)

Misra, Rajeev; Stikeleather, Ryan; Gabriele, Rebecca

2015-03-13

Assembly of the β-barrel outer membrane proteins (OMPs) is an essential cellular process in Gram-negative bacteria and in the mitochondria and chloroplasts of eukaryotes--two organelles of bacterial origin. Central to this process is the conserved β-barrel OMP that belongs to the Omp85 superfamily. In Escherichia coli, BamA is the core β-barrel OMP and, together with four outer membrane lipoproteins, BamBCDE, constitutes the β-barrel assembly machine (BAM). In this paper, we investigated the roles of BamD, an essential lipoprotein, and BamB in BamA biogenesis. Depletion of BamD caused impairment in BamA biogenesis and cessation of cell growth. These defects of BamD depletion were partly reversed by single-amino-acid substitutions mapping within the β-barrel domain of BamA. However, in the absence of BamB, the positive effects of the β-barrel substitutions on BamA biogenesis under BamD depletion conditions were nullified. By employing a BamA protein bearing one such substitution, F474L, it was demonstrated that the mutant BamA protein could not only assemble without BamD but also facilitate the assembly of wild-type BamA expressed in trans. Based on these data, we propose a model in which the Bam lipoproteins, which are localized to the outer membrane by the BAM-independent Lol pathway, aid in the creation of new BAM complexes by serving as outer membrane receptors and folding factors for nascent BamA molecules. The newly assembled BAM holocomplex then catalyzes the assembly of substrate OMPs and BamA. These in vivo findings are corroborated by recently published in vitro data. Copyright © 2014 Elsevier Ltd. All rights reserved.

Manufacture of core sub-assemblies and fertile fuel assemblies for Indian fast breeder programme

International Nuclear Information System (INIS)

Jayaraj, R.N.

2009-01-01

sintered ThO 2 . Different varieties of stainless steel are employed for the manufacture of intricate components required for Fast Breeder sub-assemblies which are not easily machinable. Large number of precision machined components are fabricated through specialized machining, forming and welding techniques and finally assembled with the help of special jigs and fixtures. Several fabrication techniques were developed like Clad Tube Crimping, Button Forming of Hexagonal Tube, Bead Forming of Spacer Wire, Welding of Clad Tubes to End Plugs and Hexagonal Tube to Foot and Handling Head. Specialized Joining Techniques like Pulsed Current GTAW are employed for the fabrication of thin- walled Fuel Elements. Developmental works are also undertaken for standardizing manufacturing techniques for Oxide Dispersion Strengthened (ODS) alloys for clad tubes of Fast Breeder Reactors, which will have an edge over conventional materials with respect to excellent resistance to void swelling and irradiation embrittlement and also capable of operating under severe conditions for extended periods. The paper highlights various developmental activities carried out for the manufacture of core sub-assemblies for the Fast Breeder Test Reactor (FBTR) and the forthcoming Prototype Fast Breeder Reactor (PFBR). (author)

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)

The characters of self-assembly core/shell nanoparticles of amphiphilic hyperbranched polyethers as drug carriers

International Nuclear Information System (INIS)

Ajun Wan; Yuxia, Kou

2008-01-01

The characters of self-assembly core/shell nanoparticles of amphiphilic hyperbranched polyethers (HP-g-PEO) as drug carriers were investigated. The HP-g-PEO consisting of hydrophobic HP-g-PEO core and hydrophilic poly(ethylene glycol) arms was prepared by the cation ring-opening polymerization. A series of HP-g-PEO samples with different degree of branching (DB) were synthesized under various reaction temperatures. Nanoparticles (NP) were obtained by self-assembly of HP-g-PEO in aqueous media. The structure of resulting HP-g-PEO was characterized by IR, 13 CNMR and GPC. Dynamic light scattering and transmission electron microscopy were applied to characterize the sizes and size distributions of NP. The results demonstrated that the mean diameters of NP were less than 100 nm, which exhibited uniform spherical formations and narrow size distributions. Using hydrophobic drug Probucol (PRO) as model drug, the particle sizes of drug loaded NP were larger than relative blank NP. The drug loading efficiency (LE) and incorporation efficiency (IE) of these NP were achieved to 35 and 89%, respectively. The in vitro release of PRO from the NP exhibited a sustained release and the cumulative drugs released for more than 600 h. The most important factor to affect drug release was the value of DB of HP-g-PEO. With the DB of HP-g-PEO increasing, the size and size distribution of NP decreased as well as the release rate. However, the small DB was beneficial to the LE of NP. Nanoparticle size and size distribution, LE, IE, and drug release rate were slightly affected by the initial solution concentration of polyethers. The co-incorporated hydrophilic drug had influence slightly on the release of drug from drug loaded NP. The results of in vitro drug release suggested that the core/shell NP performed good controlled release behaviors with potential practice as novelty drug delivery vehicles

In-core fuel management for nuclear reactor

International Nuclear Information System (INIS)

Ross, M.F.; Visner, S.

1986-01-01

This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

Controlled-release and preserved bioactivity of proteins from (self-assembled core-shell double-walled microspheres

Directory of Open Access Journals (Sweden)

Yuan W

2012-01-01

Full Text Available Weien Yuan1,2, Zhenguo Liu11Department of Neurology, Xinhua Hospital, affiliated to Shanghai Jiao Tong University School of Medicine, Shanghai, 2School of Pharmacy, Shanghai Jiao Tong University, Shanghai, People’s Republic of ChinaAbstract: In order to address preserved protein bioactivities and protein sustained-release problems, a method for preparing double-walled microspheres with a core (protein-loaded nanoparticles with a polymer-suspended granule system-formed core and a second shell (a polymer-formed shell for controlled drug release and preserved protein bioactivities has been developed using (solid-in-oil phase-in-hydrophilic oil-in-water (S/O/Oh/W phases. The method, based on our previous microsphere preparation method (solid-in-oil phase-in-hydrophilic oil-in-water (S/O/Oh/W, employs different concentric poly(D,L-lactide-co-glycolide, poly(D,L-lactide, and protein-loaded nanoparticles to produce a suspended liquid which then self-assembles to form shell-core microspheres in the hydrophilic oil phase, which are then solidified in the water phase. Variations in the preparation parameters allowed complete encapsulation by the shell phase, including the efficient formation of a poly(D,L-lactide shell encapsulating a protein-loaded nanoparticle-based poly(D,L-lactide-co-glycolide core. This method produces core-shell double-walled microspheres that show controlled protein release and preserved protein bioactivities for 60 days. Based upon these results, we concluded that the core-shell double-walled microspheres might be applied for tissue engineering and therapy for chronic diseases, etc.Keywords: protein delivery, protein stability, core-shell microspheres, dextran nanoparticles

KNK II, third core: Design Report of the breeder assemblies NR-230R, NR-231R, NR-233D and NR-234D

International Nuclear Information System (INIS)

Ricken, E.

1988-07-01

The report describes the blanket assemblies of the third core of KNK II and their operational behavior. The assemblies and their individual components are described and the design criteria, design methods and design results are presented. With the help of enveloping proofs the respectation of the design targets up to the foreseen residence time of 720 equivalent full-power days is evidenced [de

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

2007-10-01

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of

The influence of ligand charge and length on the assembly of Brome mosaic virus derived virus-like particles with magnetic core

Science.gov (United States)

Mieloch, Adam A.; Krecisz, Monika; Rybka, Jakub D.; Strugała, Aleksander; Krupiński, Michał; Urbanowicz, Anna; Kozak, Maciej; Skalski, Bohdan; Figlerowicz, Marek; Giersig, Michael

2018-03-01

Virus-like particles (VLPs) have sparked a great interest in the field of nanobiotechnology and nanomedicine. The introduction of superparamagnetic nanoparticles (SPIONs) as a core, provides potential use of VLPs in the hyperthermia therapy, MRI contrast agents and magnetically-powered delivery agents. Magnetite NPs also provide a significant improvement in terms of VLPs stability. Moreover employing viral structural proteins as self-assembling units has opened a new paths for targeted therapy, drug delivery systems, vaccines design, and many more. In many cases, the self-assembly of a virus strongly depends on electrostatic interactions between positively charged groups of the capsid proteins and negatively charged nucleic acid. This phenomenon imposes the negative net charge as a key requirement for the core nanoparticle. In our experiments, Brome mosaic virus (BMV) capsid proteins isolated from infected plants Hordeum vulgare were used. Superparamagnetic iron oxide nanoparticles (Fe3O4) with 15 nm in diameter were synthesized by thermal decomposition and functionalized with COOH-PEG-PL polymer or dihexadecylphosphate (DHP) in order to provide water solubility and negative charge required for the assembly. Nanoparticles were characterized by Transmission Electron Microscopy (TEM), Dynamic Light Scattering (DLS), Zeta Potential, Fourier Transformed Infrared Spectroscopy (FTIR) and Superconducting Quantum Interference Device (SQUID) magnetometry. TEM and DLS study were conducted to verify VLPs creation. This study demonstrates that the increase of negative surface charge is not a sufficient factor determining successful assembly. Additional steric interactions provided by longer ligands are crucial for the assembly of BMV SPION VLPs and may enhance the colloidal stability.

The influence of ligand charge and length on the assembly of Brome mosaic virus derived virus-like particles with magnetic core

Directory of Open Access Journals (Sweden)

Adam A. Mieloch

2018-03-01

Full Text Available Virus-like particles (VLPs have sparked a great interest in the field of nanobiotechnology and nanomedicine. The introduction of superparamagnetic nanoparticles (SPIONs as a core, provides potential use of VLPs in the hyperthermia therapy, MRI contrast agents and magnetically-powered delivery agents. Magnetite NPs also provide a significant improvement in terms of VLPs stability. Moreover employing viral structural proteins as self-assembling units has opened a new paths for targeted therapy, drug delivery systems, vaccines design, and many more. In many cases, the self-assembly of a virus strongly depends on electrostatic interactions between positively charged groups of the capsid proteins and negatively charged nucleic acid. This phenomenon imposes the negative net charge as a key requirement for the core nanoparticle. In our experiments, Brome mosaic virus (BMV capsid proteins isolated from infected plants Hordeum vulgare were used. Superparamagnetic iron oxide nanoparticles (Fe3O4 with 15 nm in diameter were synthesized by thermal decomposition and functionalized with COOH-PEG-PL polymer or dihexadecylphosphate (DHP in order to provide water solubility and negative charge required for the assembly. Nanoparticles were characterized by Transmission Electron Microscopy (TEM, Dynamic Light Scattering (DLS, Zeta Potential, Fourier Transformed Infrared Spectroscopy (FTIR and Superconducting Quantum Interference Device (SQUID magnetometry. TEM and DLS study were conducted to verify VLPs creation. This study demonstrates that the increase of negative surface charge is not a sufficient factor determining successful assembly. Additional steric interactions provided by longer ligands are crucial for the assembly of BMV SPION VLPs and may enhance the colloidal stability.

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.)

Chlorine-induced assembly of a cationic coordination cage with a μ5-carbonato-bridged Mn(II)24 core.

Science.gov (United States)

Xiong, Ke-Cai; Jiang, Fei-Long; Gai, Yan-Li; Yuan, Da-Qiang; Han, Dong; Ma, Jie; Zhang, Shu-Quan; Hong, Mao-Chun

2012-04-27

Chlorine caged in! The chlorine-induced assembly of six shuttlecock-like tetranuclear Mn(II) building blocks generated in situ based on p-tert-butylthiacalix[4]arene and facial anions gave rise to a novel truncated distorted octahedral cationic coordination cage with a μ(5)-carbonato-bridged Mn(II)(24) core. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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

Construction and utilization of linear empirical core models for PWR in-core fuel management

International Nuclear Information System (INIS)

Okafor, K.C.

1988-01-01

An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k ∞ profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results

Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

International Nuclear Information System (INIS)

Pavlovitchev, A.M.

2000-01-01

The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

Feasibility study of application of ductless fuel assembly to FBR

International Nuclear Information System (INIS)

Itoh, K.; Shibahara, I.

1996-01-01

Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors

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.

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.

A solution for automatic parallelization of sequential assembly code

Directory of Open Access Journals (Sweden)

Kovačević Đorđe

2013-01-01

Full Text Available Since modern multicore processors can execute existing sequential programs only on a single core, there is a strong need for automatic parallelization of program code. Relying on existing algorithms, this paper describes one new software solution tool for parallelization of sequential assembly code. The main goal of this paper is to develop the parallelizator which reads sequential assembler code and at the output provides parallelized code for MIPS processor with multiple cores. The idea is the following: the parser translates assembler input file to program objects suitable for further processing. After that the static single assignment is done. Based on the data flow graph, the parallelization algorithm separates instructions on different cores. Once sequential code is parallelized by the parallelization algorithm, registers are allocated with the algorithm for linear allocation, and the result at the end of the program is distributed assembler code on each of the cores. In the paper we evaluate the speedup of the matrix multiplication example, which was processed by the parallelizator of assembly code. The result is almost linear speedup of code execution, which increases with the number of cores. The speed up on the two cores is 1.99, while on 16 cores the speed up is 13.88.

A study on the core characteristics of the fact breed reactor

International Nuclear Information System (INIS)

Cho, M.; Ryu, K.G.; Soh, D.S.; Kim, Y.C.; Lee, K.W.; Kim, Y.I.; Lim, J.C.; Oh, K.B.

1983-01-01

In order to investigate equilibrium core characteristics of a large LMFBR, nuclear-thermal hydraulic and safety-related characteristics are considered for the equilbrium core of Super-Phenix 1. Using the nuclear computational system for a large LMFBR (KAERI-26G cross section library/1DX/2DB), bias factor of the effective multiplication factor, refueling pattern and enrichments of refueled fuel assemblies are determined. Nuclear characteristics such as criticality, power distribution, and breeding gain are also obtained for the initial core to the equilibrium core. Based on the assembly power distribution of the equilibrium core, coolant flow distributions are determined with the use of THI3D code. Temperature fields within the core assemblies and pressure drop across the core assemblies are also analyzed. In addition, the subchannel analysis of the hottest assembly under the steady state is performed with COBRA-IV-I code. Finally, transient behaviors of reactivity, power, and temperature due to the loss of flow accident without scram are considered and the boiling initiation time of coolant in the fuel assembly is determined for the equilibrium core with SACO code. (Author)

Evaluation of core distortion in FBR

International Nuclear Information System (INIS)

Ikarimoto, I.; Tanaka, M.; Okubo, Y.

1984-01-01

The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

Self-assembly of core-polyethylene glycol-lipid shell (CPLS) nanoparticles and their potential as drug delivery vehicles

Science.gov (United States)

Shen, Zhiqiang; Loe, David T.; Awino, Joseph K.; Kröger, Martin; Rouge, Jessica L.; Li, Ying

2016-08-01

Herein a new multifunctional formulation, referred to as a core-polyethylene glycol-lipid shell (CPLS) nanoparticle, has been proposed and studied in silico via large scale coarse-grained molecular dynamics simulations. A PEGylated core with surface tethered polyethylene glycol (PEG) chains is used as the starting configuration, where the free ends of the PEG chains are covalently bonded with lipid molecules (lipid heads). A complete lipid bilayer is formed at the surface of the PEGylated particle core upon addition of free lipids, driven by the hydrophobic properties of the lipid tails, leading to the formation of a CPLS nanoparticle. The self-assembly process is found to be sensitive to the grafting density and molecular weight of the tethered PEG chains, as well as the amount of free lipids added. At low grafting densities the assembly of CPLS nanoparticles cannot be accomplished. As demonstrated by simulations, a lipid bud/vesicle can be formed on the surface when an excess amount of free lipids is added at high grafting density. Therefore, the CPLS nanoparticles can only be formed under appropriate conditions of both PEG and free lipids. The CPLS nanoparticle has been recognized to be able to store a large quantity of water molecules, particularly with high molecular weight of PEG chains, indicating its capacity for carrying hydrophilic molecules such as therapeutic biomolecules or imaging agents. Under identical size and surface chemistry conditions of a liposome, it has been observed that the CPLS particle can be more efficiently wrapped by the lipid membrane, indicating its potential for a greater efficiency in delivering its hydrophilic cargo. As a proof-of-concept, the experimental realization of CPLS nanoparticles is explicitly demonstrated in this study. To test the capacity of the CPLS to store small molecule cargo a hydrophilic dye was successfully encapsulated in the particles' water soluble layer. The results of this study show the power and

HANARO core channel flow-rate measurement

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

Initial charge reactor core

International Nuclear Information System (INIS)

Kiyono, Takeshi

1984-01-01

Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

Role of ligand-ligand vs. core-core interactions in gold nanoclusters.

Science.gov (United States)

Milowska, Karolina Z; Stolarczyk, Jacek K

2016-05-14

The controlled assembly of ligand-coated gold nanoclusters (NCs) into larger structures paves the way for new applications ranging from electronics to nanomedicine. Here, we demonstrate through rigorous density functional theory (DFT) calculations employing novel functionals accounting for van der Waals forces that the ligand-ligand interactions determine whether stable assemblies can be formed. The study of NCs with different core sizes, symmetry forms, ligand lengths, mutual crystal orientations, and in the presence of a solvent suggests that core-to-core van der Waals interactions play a lesser role in the assembly. The dominant interactions originate from combination of steric effects, augmented by ligand bundling on NC facets, and related to them changes in electronic properties induced by neighbouring NCs. We also show that, in contrast to standard colloidal theory approach, DFT correctly reproduces the surprising experimental trends in the strength of the inter-particle interaction observed when varying the length of the ligands. The results underpin the importance of understanding NC interactions in designing gold NCs for a specific function.

Evidence that assembly of the yeast cytochrome bc1 complex involves formation of a large core structure in the inner mitochondrial membrane

Science.gov (United States)

Zara, Vincenzo; Conte, Laura; Trumpower, Bernard L.

2009-01-01

The assembly status of the cytochrome bc1 complex has been analyzed in distinct yeast deletion strains in which genes for one or more of the bc1 subunits had been deleted. In all the yeast strains tested a bc1 sub-complex of about 500 kDa was found when the mitochondrial membranes were analyzed by blue native electrophoresis. The subsequent molecular characterization of this sub-complex, carried out in the second dimension by SDS-PAGE and immunodecoration, revealed the presence of the two catalytic subunits cytochrome b and cytochrome c1, associated with the non catalytic subunits core protein 1, core protein 2, Qcr7p and Qcr8p. Altogether these bc1 subunits build up the core structure of the cytochrome bc1 complex which is then able to sequentially bind the remaining subunits, such as Qcr6p, Qcr9p, the Rieske iron-sulfur protein and Qcr10p. This bc1 core structure may represent a true assembly intermediate during the maturation of the bc1 complex, first because of its wide distribution in distinct yeast deletion strains and second for its characteristics of stability which resemble those of the intact homodimeric bc1 complex. Differently from this latter, however, the bc1 core structure is not able to interact with the cytochrome c oxidase complex to form respiratory supercomplexes. The characterization of this novel core structure of the bc1 complex provides a number of new elements for clarification of the molecular events leading to the maturation of the yeast cytochrome bc1 complex in the inner mitochondrial membrane. PMID:19236481

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Decay heat and activity of the structural materials of the fuel and blanket assemblies of the second and third core of KNK II

International Nuclear Information System (INIS)

Winterhagen, D.

1986-06-01

The decay heat and activity caused by structural materials have been calculated for the fuel assemblies of KNK II (second and third core) with a residence time of 720 equivalent full-power days (efpd) and the blanket assemblies with 1880 efpd. The values are given for the different zones of the assemblies (head, active zone, fission gas plenum, foot and stellite area) for decay times from 1 to 20 years. For decay times beyond 2 years more than 80 % of the decay heat are caused by the Co60-decay, more than 60 % of which result from the stellite in the foot area [de

DARK MATTER CORES IN THE FORNAX AND SCULPTOR DWARF GALAXIES: JOINING HALO ASSEMBLY AND DETAILED STAR FORMATION HISTORIES

International Nuclear Information System (INIS)

Amorisco, N. C.; Zavala, J.; De Boer, T. J. L.

2014-01-01

We combine the detailed star formation histories of the Fornax and Sculptor dwarf spheroidals with the mass assembly history of their dark matter (DM) halo progenitors to estimate if the energy deposited by Type II supernovae (SNe II) is sufficient to create a substantial DM core. Assuming the efficiency of energy injection of the SNe II into DM particles is ε gc = 0.05, we find that a single early episode, z ≳ z infall , that combines the energy of all SNe II due to explode over 0.5 Gyr is sufficient to create a core of several hundred parsecs in both Sculptor and Fornax. Therefore, our results suggest that it is energetically plausible to form cores in cold dark matter (CDM) halos via early episodic gas outflows triggered by SNe II. Furthermore, based on CDM merger rates and phase-space density considerations, we argue that the probability of a subsequent complete regeneration of the cusp is small for a substantial fraction of dwarf-size halos

Neutronic Design of KALIMER-600 Core with Moderator Rods

International Nuclear Information System (INIS)

Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

2004-01-01

Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

Preliminary study on the feasibility of ductless fuel assembly for fast reactors

International Nuclear Information System (INIS)

Shibahara, Itaru; Enokido, Yuji

1988-01-01

Preliminary study on the feasibility of ductless fuel assembly for fast reactors has been conducted. The primary concern is with forecasting the thermal hydraulic characteristics and the heat removal efficiency from the core. The thermal hydraulic analysis revealed the coolant mixing in the core at steady state operating condition was not intensive and the coolant temperature increase was almost proportional to the power of each assembly. The hot spot analysis of the ductless core indicated that the hottest temperature in the core could be comparable with the temperature of the conventional ducted core, even in case the radial power flattening was not actively pursued but with adopting ducted radial blanket assemblies. Under off-normal conditions, the ductless core had improved heat removal capability which was caused by inter-assembly coolant flow. The study has indicated the feasibility of the ductless fuel assembly for fast reactors. The experiments to demonstrate the feasibility will be the next key process for the development. (author)

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

Classification of functioning and impairment: the development of ICF core sets for autism spectrum disorder.

Science.gov (United States)

Bölte, Sven; de Schipper, Elles; Robison, John E; Wong, Virginia C N; Selb, Melissa; Singhal, Nidhi; de Vries, Petrus J; Zwaigenbaum, Lonnie

2014-02-01

Given the variability seen in Autism Spectrum Disorder (ASD), accurate quantification of functioning is vital to studying outcome and quality of life in affected individuals. The International Classification of Functioning, Disability and Health (ICF) provides a comprehensive, universally accepted framework for the description of health-related functioning. ICF Core Sets are shortlists of ICF categories that are selected to capture those aspects of functioning that are most relevant when describing a person with a specific condition. In this paper, the authors preview the process for developing ICF Core Sets for ASD, a collaboration with the World Health Organization and the ICF Research Branch. The ICF Children and Youth version (ICF-CY) was derived from the ICF and designed to capture the specific situation of the developing child. As ASD affects individuals throughout the life span, and the ICF-CY includes all ICF categories, the ICF-CY will be used in this project ("ICF(-CY)" from now on). The ICF(-CY) categories to be included in the ICF Core Sets for ASD will be determined at an ICF Core Set Consensus Conference, where evidence from four preparatory studies (a systematic review, an expert survey, a patient and caregiver qualitative study, and a clinical cross-sectional study) will be integrated. Comprehensive and Brief ICF Core Sets for ASD will be developed with the goal of providing useful standards for research and clinical practice and generating a common language for functioning and impairment in ASD in different areas of life and across the life span. © 2013 International Society for Autism Research, Wiley Periodicals, Inc.

Construction of the HTTR in-core components

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

1996-01-01

The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

Fuel assembly

International Nuclear Information System (INIS)

Kurihara, Kunitoshi; Azekura, Kazuo.

1992-01-01

In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

Holddown device for nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1978-01-01

An apparatus for preventing ''floating'' of nuclear-reactor fuel assemblies due to hydraulic forces is disclosed. The apparatus uses a holddown column made of the same material as the core barrel. The column is positioned in a center guide-tube location in the fuel assembly in such a manner as to enable it either to slide within the center guide tube or, if the center guide tube is replaced by the column, to slide through openings in the spacer grids. The lower end of the holddown column engages the lower end fitting of the fuel assembly, and the upper end of the column engages a flow plate to which holddown force is applied. As a consequence of this arrangement, holddown force is transmitted from the flow plate through the holddown column to the lower end fitting. Movement of the fuel assembly is thereby prevented without a compression load being applied to the fuel-assemb1ly structure. In addition, variations due to thermal expansion in the distance between the lower core plate and the upper core plate are largely made up for by corresponding variations in the holddown column because the holddown column and the core barrel can be made of the same material

Thermal margin model for transition core of KSNP

International Nuclear Information System (INIS)

Nahm, Kee Yil; Lim, Jong Seon; Park, Sung Kew; Chun, Chong Kuk; Hwang, Sun Tack

2004-01-01

The PLUS7 fuel was developed with mixing vane grids for KSNP. For the transition core partly loaded with the PLUS7 fuels, the procedure to set up the optimum thermal margin model of the transition core was suggested by introducing AOPM concept into the screening method which determines the limiting assembly. According to the procedure, the optimum thermal margin model of the first transition core was set up by using a part of nuclear data for the first transition and the homogeneous core with PLUS7 fuels. The generic thermal margin model of PLUS7 fuel was generated with the AOPM of 138%. The overpower penalties on the first transition core were calculated to be 1.0 and 0.98 on the limiting assembly and the generic thermal margin model, respectively. It is not usual case to impose the overpower penalty on reload cores. It is considered that the lack of channel flow due to the difference of pressure drop between PLUS7 and STD fuels results in the decrease of DNBR. The AOPM of the first transition core is evaluated to be about 135% by using the optimum generic thermal margin model which involves the generic thermal margin model and the total overpower penalty. The STD fuel is not included among limiting assembly candidates in the second transition core, because they have much lower pin power than PLUS7 fuels. The reduced number of STD fuels near the limiting assembly candidates the flow from the limiting assembly to increase the thermal margin for the second transition core. It is expected that cycle specific overpower penalties increase the thermal margin for the transition core. Using the procedure to set up the optimum thermal margin model makes sure that the enhanced thermal margin of PLUS7 fuel can be sufficiently applied to not only the homogeneous core but also the transition core

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-09-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

Hepatitis B Virus Core Protein Dephosphorylation Occurs during Pregenomic RNA Encapsidation.

Science.gov (United States)

Zhao, Qiong; Hu, Zhanying; Cheng, Junjun; Wu, Shuo; Luo, Yue; Chang, Jinhong; Hu, Jianming; Guo, Ju-Tao

2018-07-01

Hepatitis B virus (HBV) core protein consists of an N-terminal assembly domain and a C-terminal domain (CTD) with seven conserved serines or threonines that are dynamically phosphorylated/dephosphorylated during the viral replication cycle. Sulfamoylbenzamide derivatives are small molecular core protein allosteric modulators (CpAMs) that bind to the heteroaryldihydropyrimidine (HAP) pocket between the core protein dimer-dimer interfaces. CpAM binding alters the kinetics and pathway of capsid assembly and can result in the formation of morphologically "normal" capsids devoid of viral pregenomic RNA (pgRNA) and DNA polymerase. In order to investigate the mechanism underlying CpAM inhibition of pgRNA encapsidation, we developed an immunoblotting assay that can resolve core protein based on its phosphorylation status and demonstrated, for the first time, that core protein is hyperphosphorylated in free dimers and empty capsids from both mock-treated and CpAM-treated cells but is hypophosphorylated in pgRNA- and DNA-containing nucleocapsids. Interestingly, inhibition of pgRNA encapsidation by a heat shock protein 90 (HSP90) inhibitor prevented core protein dephosphorylation. Moreover, core proteins with point mutations at the wall of the HAP pocket, V124A and V124W, assembled empty capsids and nucleocapsids with altered phosphorylation status. The results thus suggest that core protein dephosphorylation occurs in the assembly of pgRNA and that interference with the interaction between core protein subunits at dimer-dimer interfaces during nucleocapsid assembly alters not only capsid structure, but also core protein dephosphorylation. Hence, inhibition of pgRNA encapsidation by CpAMs might be due to disruption of core protein dephosphorylation during nucleocapsid assembly. IMPORTANCE Dynamic phosphorylation of HBV core protein regulates multiple steps of viral replication. However, the regulatory function was mainly investigated by phosphomimetic mutagenesis, which

Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

International Nuclear Information System (INIS)

Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

2003-01-01

The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

Nuclear reactor core flow baffling

International Nuclear Information System (INIS)

Berringer, R.T.

1979-01-01

A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

Evidence that the assembly of the yeast cytochrome bc1 complex involves the formation of a large core structure in the inner mitochondrial membrane.

Science.gov (United States)

Zara, Vincenzo; Conte, Laura; Trumpower, Bernard L

2009-04-01

The assembly status of the cytochrome bc(1) complex has been analyzed in distinct yeast deletion strains in which genes for one or more of the bc(1) subunits were deleted. In all the yeast strains tested, a bc(1) sub-complex of approximately 500 kDa was found when the mitochondrial membranes were analyzed by blue native electrophoresis. The subsequent molecular characterization of this sub-complex, carried out in the second dimension by SDS/PAGE and immunodecoration, revealed the presence of the two catalytic subunits, cytochrome b and cytochrome c(1), associated with the noncatalytic subunits core protein 1, core protein 2, Qcr7p and Qcr8p. Together, these bc(1) subunits build up the core structure of the cytochrome bc(1) complex, which is then able to sequentially bind the remaining subunits, such as Qcr6p, Qcr9p, the Rieske iron-sulfur protein and Qcr10p. This bc(1) core structure may represent a true assembly intermediate during the maturation of the bc(1) complex; first, because of its wide distribution in distinct yeast deletion strains and, second, for its characteristics of stability, which resemble those of the intact homodimeric bc(1) complex. By contrast, the bc(1) core structure is unable to interact with the cytochrome c oxidase complex to form respiratory supercomplexes. The characterization of this novel core structure of the bc(1) complex provides a number of new elements clarifying the molecular events leading to the maturation of the yeast cytochrome bc(1) complex in the inner mitochondrial membrane.

Proposal for the award of a contract for the supply of moulded assemblies of tape-wound cut cores of grain-oriented silicon-steel for the LHC beam dumping system

CERN Document Server

2001-01-01

This document concerns the award of a contract for the supply of moulded assemblies of tape-wound cut cores of grain-oriented silicon-steel for the three different types of fast pulsed magnets for the beam dumping system of the LHC. Following a market survey carried out among 14 firms in four Member States and six firms in two non-Member States, a call for tenders (IT-2674/SL/LHC) was sent on 26 January 2001 to one firm in one Member State. By the closing date, CERN had received a tender from the firm. The Finance Committee is invited to agree to the negotiation of a contract with TELMAG MAGNETIC COMPONENTS (UK), the only bidder, for the supply of 2885 moulded assemblies of tape-wound cut cores of grain-oriented silicon-steel, of three different types, for a total amount of 2 414 664 euros (3 692 147 Swiss francs), not subject to revision, with options for 560 additional moulded core assemblies of three different types, for an additional amount of 468 496 euros (716 355 Swiss francs), not subject to revision,...

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-01-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

Tandem fusion of hepatitis B core antigen allows assembly of virus-like particles in bacteria and plants with enhanced capacity to accommodate foreign proteins.

Directory of Open Access Journals (Sweden)

Hadrien Peyret

Full Text Available The core protein of the hepatitis B virus, HBcAg, assembles into highly immunogenic virus-like particles (HBc VLPs when expressed in a variety of heterologous systems. Specifically, the major insertion region (MIR on the HBcAg protein allows the insertion of foreign sequences, which are then exposed on the tips of surface spike structures on the outside of the assembled particle. Here, we present a novel strategy which aids the display of whole proteins on the surface of HBc particles. This strategy, named tandem core, is based on the production of the HBcAg dimer as a single polypeptide chain by tandem fusion of two HBcAg open reading frames. This allows the insertion of large heterologous sequences in only one of the two MIRs in each spike, without compromising VLP formation. We present the use of tandem core technology in both plant and bacterial expression systems. The results show that tandem core particles can be produced with unmodified MIRs, or with one MIR in each tandem dimer modified to contain the entire sequence of GFP or of a camelid nanobody. Both inserted proteins are correctly folded and the nanobody fused to the surface of the tandem core particle (which we name tandibody retains the ability to bind to its cognate antigen. This technology paves the way for the display of natively folded proteins on the surface of HBc particles either through direct fusion or through non-covalent attachment via a nanobody.

Cytotoxicity of Paclitaxel in biodegradable self-assembled core-shell poly(lactide-co-glycolide ethylene oxide fumarate) nanoparticles.

Science.gov (United States)

He, Xuezhong; Ma, Junyu; Mercado, Angel E; Xu, Weijie; Jabbari, Esmaiel

2008-07-01

Biodegradable core-shell polymeric nanoparticles (NPs), with a hydrophobic core and hydrophilic shell, are developed for surfactant-free encapsulation and delivery of Paclitaxel to tumor cells. Poly (lactide-co-glycolide fumarate) (PLGF) and Poly (lactide-fumarate) (PLAF) were synthesized by condensation polymerization of ultra-low molecular weight poly(L: -lactide-co-glycolide) (ULMW PLGA) with fumaryl chloride (FuCl). Similarly, poly(lactide-co-ethylene oxide fumarate) (PLEOF) macromer was synthesized by reacting ultra-low molecular weight poly(L: -lactide) (ULMW PLA) and PEG with FuCl. The blend PLGF/PLEOF and PLAF/PLEOF macromers were self-assembled into NPs by dialysis. The NPs were characterized with respect to particle size distribution, morphology, and loading efficiency. The physical state and miscibility of Paclitaxel in NPs were characterized by differential scanning calorimetry. Tumor cell uptake and cytotoxicity of Paclitaxel loaded NPs were measured by incubation with HCT116 human colon carcinoma cells. The distribution of NPs in vivo was assessed with Apc(Min/+)mouse using infrared imaging. PLEOF macromer, due to its amphiphilic nature, acted as a surface active agent in the process of self-assembly which produced core-shell NPs with PLGF/PLAF and PLEOF macromers as the core and shell, respectively. The encapsulation efficiency ranged from 70 to 56% and it was independent of the macromer but decreased with increasing concentration of Paclitaxel. Most of the PLGF and PLAF NPs degraded in 15 and 28 days, respectively, which demonstrated that the release was dominated by hydrolytic degradation and erosion of the matrix. As the concentration of Paclitaxel was increased from 0 to 10, and 40 mug/ml, the viability of HCT116 cells incubated with free Paclitaxel decreased from 100 to 65 and 40%, respectively, while those encapsulated in PLGF/PLEOF NPs decreased from 93 to 54 and 28%. Groups with Paclitaxel loaded NPs had higher cytotoxicity compared to

Environment-based pin-power reconstruction method for homogeneous core calculations

International Nuclear Information System (INIS)

Leroyer, H.; Brosselard, C.; Girardi, E.

2012-01-01

Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

Calculation of the ex-core neutron noise induced by fuel vibrations in PWRs

International Nuclear Information System (INIS)

Tran Hoai Nam; Cao Van Chung; Hoang Thanh Phi Hung; Hoang Van Khanh

2015-01-01

Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor (PWR) cores has been performed to investigate the effect of cycle burnup on the properties of the ex-core detector noise. Pendular vibrations of individual fuel assemblies were assumed to occur at different locations in the core. The auto power spectra density (APSD) of the ex-core detector noise was evaluated with the assumption of stochastic vibrations along a random two-dimensional trajectory. The results show that no general monotonic variation of APSD was found. The increase of APSD occurs predominantly for peripheral assemblies. Assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with the more realistic perturbation model, the effect of the peripheral assemblies will dominate and the increase of the amplitude of the ex-core neutron noise with burnup can be confirmed. (author)

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

International Nuclear Information System (INIS)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

2016-01-01

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

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)

Core subjects at the end of primary school: identifying and explaining relative strengths of children with specific language impairment (SLI)

Science.gov (United States)

Durkin, Kevin; Mok, Pearl L H; Conti-Ramsden, Gina

2015-01-01

Background In general, children with specific language impairment (SLI) tend to fall behind their typically developing (TD) peers in educational attainment. Less is known about how children with SLI fare in particular areas of the curriculum and what predicts their levels of performance. Aims To compare the distributions of performance of children with SLI in three core school subjects (English, Mathematics and Science); to test the possibility that performance would vary across the core subjects; and to examine the extent to which language impairment predicts performance. Methods & Procedures This study was conducted in England and reports historical data on educational attainments. Teacher assessment and test scores of 176 eleven-year-old children with SLI were examined in the three core subjects and compared with known national norms. Possible predictors of performance were measured, including language ability at ages 7 and 11, educational placement type, and performance IQ. Outcomes & Results Children with SLI, compared with national norms, were found to be at a disadvantage in core school subjects. Nevertheless, some children attained the levels expected of TD peers. Performance was poorest in English; relative strengths were indicated in Science and, to a lesser extent, in Mathematics. Language skills were significant predictors of performance in all three core subjects. PIQ was the strongest predictor for Mathematics. For Science, both early language skills at 7 years and PIQ made significant contributions. Conclusions & Implications Language impacts on the school performance of children with SLI, but differentially across subjects. English for these children is the most challenging of the core subjects, reflecting the high levels of language demand it incurs. Science is an area of relative strength and mathematics appears to be intermediate, arguably because some tasks in these subjects can be performed with less reliance on verbal processing. Many children

Core subjects at the end of primary school: identifying and explaining relative strengths of children with specific language impairment (SLI).

Science.gov (United States)

Durkin, Kevin; Mok, Pearl L H; Conti-Ramsden, Gina

2015-01-01

In general, children with specific language impairment (SLI) tend to fall behind their typically developing (TD) peers in educational attainment. Less is known about how children with SLI fare in particular areas of the curriculum and what predicts their levels of performance. To compare the distributions of performance of children with SLI in three core school subjects (English, Mathematics and Science); to test the possibility that performance would vary across the core subjects; and to examine the extent to which language impairment predicts performance. This study was conducted in England and reports historical data on educational attainments. Teacher assessment and test scores of 176 eleven-year-old children with SLI were examined in the three core subjects and compared with known national norms. Possible predictors of performance were measured, including language ability at ages 7 and 11, educational placement type, and performance IQ. Children with SLI, compared with national norms, were found to be at a disadvantage in core school subjects. Nevertheless, some children attained the levels expected of TD peers. Performance was poorest in English; relative strengths were indicated in Science and, to a lesser extent, in Mathematics. Language skills were significant predictors of performance in all three core subjects. PIQ was the strongest predictor for Mathematics. For Science, both early language skills at 7 years and PIQ made significant contributions. Language impacts on the school performance of children with SLI, but differentially across subjects. English for these children is the most challenging of the core subjects, reflecting the high levels of language demand it incurs. Science is an area of relative strength and mathematics appears to be intermediate, arguably because some tasks in these subjects can be performed with less reliance on verbal processing. Many children with SLI do have the potential to reach or exceed educational targets that are set

SWAP-Assembler 2: Optimization of De Novo Genome Assembler at Large Scale

Energy Technology Data Exchange (ETDEWEB)

Meng, Jintao; Seo, Sangmin; Balaji, Pavan; Wei, Yanjie; Wang, Bingqiang; Feng, Shengzhong

2016-08-16

In this paper, we analyze and optimize the most time-consuming steps of the SWAP-Assembler, a parallel genome assembler, so that it can scale to a large number of cores for huge genomes with the size of sequencing data ranging from terabyes to petabytes. According to the performance analysis results, the most time-consuming steps are input parallelization, k-mer graph construction, and graph simplification (edge merging). For the input parallelization, the input data is divided into virtual fragments with nearly equal size, and the start position and end position of each fragment are automatically separated at the beginning of the reads. In k-mer graph construction, in order to improve the communication efficiency, the message size is kept constant between any two processes by proportionally increasing the number of nucleotides to the number of processes in the input parallelization step for each round. The memory usage is also decreased because only a small part of the input data is processed in each round. With graph simplification, the communication protocol reduces the number of communication loops from four to two loops and decreases the idle communication time. The optimized assembler is denoted as SWAP-Assembler 2 (SWAP2). In our experiments using a 1000 Genomes project dataset of 4 terabytes (the largest dataset ever used for assembling) on the supercomputer Mira, the results show that SWAP2 scales to 131,072 cores with an efficiency of 40%. We also compared our work with both the HipMER assembler and the SWAP-Assembler. On the Yanhuang dataset of 300 gigabytes, SWAP2 shows a 3X speedup and 4X better scalability compared with the HipMer assembler and is 45 times faster than the SWAP-Assembler. The SWAP2 software is available at https://sourceforge.net/projects/swapassembler.

Development of conceptual nuclear design of 10MWt research reactor core

International Nuclear Information System (INIS)

Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

2008-03-01

KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

Sub-assembly accident protection instrumentation systems

International Nuclear Information System (INIS)

Vaughan, G.J.; Lunt, A.R.W.; Evans, N.J.; Lawrence, L.A.J.

1982-01-01

The possibility of an incident in a sub-assembly progressing to the stage at which the whole core may be at hazard has to be guarded against. It is proposed that for CDFR specific instrumentation will be provided to protect against this incident. Three such systems are described, these are: Acoustic Boiling Noise Detection, Burst Pin Detection and Individual Sub-Assembly Thermocouple (ISAT) monitoring. In the ISAT case, multiplexers and microprocessors are employed, using novel techniques to ensure failure-to-safety. The role of these systems and the implementation of them in the reactor design are also considered. It is concluded that sufficient protection can be provided for both core and breeder sub-assemblies

In Vitro Assembly of the Outer Core of the Lipopolysaccharide from Escherichia coli K-12 and Salmonella typhimurium

Science.gov (United States)

2015-01-01

There are five distinct core structures in the lipopolysaccharides of Escherichia coli and at least two in Salmonella isolates, which vary principally in the outer core oligosaccharide. Six outer core glycosyltransferases, E. coli K-12 WaaG, WaaB, and WaaO and Salmonella typhimurium WaaI, WaaJ, and WaaK, were cloned, overexpressed, and purified. A novel substrate for WaaG was isolated from ΔwaaG E. coli overexpressing the lipid A phosphatase lpxE and the lipid A late acyltransferase lpxM. The action of lpxE and lpxM in the ΔwaaG background yielded heptose2-1-dephospho Kdo2-lipid A, a 1-dephosphorylated hexa-acylated lipid A with the inner core sugars that is easily isolated by organic extraction. Using this structurally defined acceptor and commercially available sugar nucleotides, each outer core glycosyltransferases was assayed in vitro. We show that WaaG and WaaB add a glucose and galactose sequentially to heptose2-1-dephospho Kdo2-lipid A. E. coli K-12 WaaO and S. typhimurium WaaI add a galactose to the WaaG/WaaB product but can also add a galactose to the WaaG product directly without the branched core sugar added by WaaB. Both WaaI and WaaO require divalent metal ions for optimal activity; however, WaaO, unlike WaaI, can add several glucose residues to its lipid acceptor. Using the product of WaaG, WaaB, and WaaI, we show that S. typhimurium WaaJ and WaaK transfer a glucose and N-acetylglucosamine, respectively, to yield the full outer core. This is the first demonstration of the in vitro assembly of the outer core of the lipopolysaccharide using defined lipid A-oligosaccharide acceptors and sugar donors. PMID:24479701

Reactor core

International Nuclear Information System (INIS)

Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

1990-01-01

Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

In-core fuel management via perturbation theory

International Nuclear Information System (INIS)

Mingle, J.O.

1975-01-01

A two-step procedure is developed for the optimization of in-core nuclear fuel management using perturbation theory to predict the effects of various core configurations. The first procedure is a cycle cost minimization using linear programming with a zoned core and discrete burnup groups. The second program utilizes an individual fuel assembly shuffling sequence to minimize the maldistribution of power generation. This latter quantity is represented by a figure of merit or by an assembly power peaking factor. A pressurized water reactor example calculation is utilized. 24 references

Study on HANARO core conversion using U-Mo fuel

International Nuclear Information System (INIS)

Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

2002-01-01

Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

Manufacturing development for the SAFE 100 kW core

International Nuclear Information System (INIS)

Carter, Robert; Roman, Jose; Salvail, Pat

2002-01-01

In stark contrast to what is sometimes considered the norm in traditional manufacturing processes, engineers at the Marshall Space Flight Center (MSFC) arc in the practice of altering the standard in an effort to realize other potential methods in core manufacturing. While remaining within the bounds of the materials database, we are researching into core manufacturing techniques that may have been overlooked in the past due to funding and/or time constraints. To augment proven core fabrication capabilities we are pursuing plating processes as another possible method for core build-up and assembly. Although brazing and a proprietary HIP cycle are used for module assembly (proven track record for stability and endurance), it is prudent to pursue secondary or backup methods of module and core assembly. For this reason heat tube manufacture and module assembly by means of plating is being investigated. Potentially, the plating processes will give engineers the ability to manufacture replacement modules for any module that might fail to perform nominally, and to assemble/disassemble a complete core in much less time than would be required for the conventional Braze-HIP process. Another area of improvement in core manufacturing capabilities is the installation of a sodium and lithium liquid metal heat pipe fill machine. This, along with the ability to Electron Beam Weld heat pipe seals and wet-in the pipes in the necessary vacuum atmosphere, will eliminate the need to ship potentially hazardous components outside for processing. In addition to developing core manufacturing techniques, the SAFE manufacturing team has been evaluating the thermal heat transfer characteristics, and manufacturability of several heat exchanger design concepts

Characterization of neutron leakage probability, k /SUB eff/ , and critical core surface mass density of small reactor assemblies through the Trombay criticality formula

International Nuclear Information System (INIS)

Kumar, A.; Rao, K.S.; Srinivasan, M.

1983-01-01

The Trombay criticality formula (TCF) has been derived by incorporating a number of well-known concepts of criticality physics to enable prediction of changes in critical size or k /SUB eff/ following alterations in geometrical and physical parameters of uniformly reflected small reactor assemblies characterized by large neutron leakage from the core. The variant parameters considered are size, shape, density and diluent concentration of the core, and density and thickness of the reflector. The effect of these changes (except core size) manifests, through sigma /SUB c/ the critical surface mass density of the ''corresponding critical core,'' that sigma, the massto-surface-area ratio of the core,'' is essentially a measure of the product /rho/ extended to nonspherical systems and plays a dominant role in the TCF. The functional dependence of k /SUB eff/ on sigma/sigma /SUB c/ , the system size relative to critical, is expressed in the TCF through two alternative representations, namely the modified Wigner rational form and, an exponential form, which is given

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

Feasibility study on nuclear core design for soluble boron free small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-04-29

A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

Recombinant in vitro assembled hepatitis C virus core particles induce strong specific immunity enhanced by formulation with an oil-based adjuvant

Directory of Open Access Journals (Sweden)

NELSON ACOSTA-RIVERO

2009-01-01

Full Text Available In the present work, immunogenicity of recombinant in vitro assembled hepatitis C virus core particles, HCcAg.120-VLPs, either alone or in combination with different adjuvants was evaluated in BALB/c mice. HCcAg.120-VLPs induced high titers of anti-HCcAg.120 antibodies and virus-specific cellular immune responses. Particularly, HCcAg.120-VLPs induced specific delayed type hypersensitivity, and generated a predominant T helper 1 cytokine pro file in immunized mice. In addition, HCcAg.120-VLPs prime splenocytes proliferate in vitro against different HCcAg.120-specific peptides, depending on either the immunization route or the adjuvant used. Remarkably, immunization with HCcAg.120-VLPs/Montanide ISA888 formulation resulted in a significant control of vaccinia virus titer in mice after challenge with a recombinant vaccinia virus expressing HCV core protein, vvCore. Animals immunized with this formulation had a marked increase in the number of IFN-γ producing spleen cells, after stimulation with P815 cells infected with vvCore. These results suggest the use of recombinant HCV core particles as components of therapeutic or preventive vaccine candidates against HCV.

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Yoon, K. H.; Lee, C. B.

2014-01-01

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

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

Economic optimization of PWR cores with ROSA

International Nuclear Information System (INIS)

Verhagen, F.C.M.; Wakker, P.H.

2005-01-01

The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

Guide tube insert assembly for use in a nuclear reactor

International Nuclear Information System (INIS)

Hopkins, R.J.; Land, J.T.

1992-01-01

This patent describes an internals assembly for a nuclear reactor of the type including an upper support plate and an upper core plate, each having apertures for conducting control rod assemblies into an out of fuel assemblies with the apertures of the upper support plate being aligned with the apertures of the upper core plate, a guide tube insert assembly comprising: an elongated tubular body extending between at least one of the aligned apertures formed in the upper support plate and the upper core plate; guide plates within the elongated tubular body, each of the guide plates having a planar surface extending substantially perpendicular to an axial direction of the tubular body; at least one interconnecting means for interconnecting the guide plates into a guide tube insert assembly such that the guide plates are simultaneously mountable within and removable from the elongated body, and the periphery of each of the guide plates is spaced apart from the inner walls of the elongated tubular body at every point when the insert assembly is mounted within the tubular body, and a stabilizing means for securing the lowermost guide plate of the guide tube insert assembly within the elongated tubular body to prevent rotational and lateral movement between the guide tube insert assembly and the tubular body

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.

A classification scheme for LWR fuel assemblies

International Nuclear Information System (INIS)

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

Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

International Nuclear Information System (INIS)

Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

1988-01-01

This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

Nuclear fuel assemblies

International Nuclear Information System (INIS)

Natori, Hisahide; Kurihara, Kunitoshi.

1982-01-01

Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

Effects of Lateral and Terminal Chains of X-Shaped Bolapolyphiles with Oligo(phenylene ethynylene Cores on Self-Assembly Behavior. Part 2: Domain Formation by Self-Assembly in Lipid Bilayer Membranes

Directory of Open Access Journals (Sweden)

Stefan Werner

2017-09-01

Full Text Available Supramolecular self-assembly of membrane constituents within a phospholipid bilayer creates complex functional platforms in biological cells that operate in intracellular signaling, trafficking and membrane remodeling. Synthetic polyphilic compounds of macromolecular or small size can be incorporated into artificial phospholipid bilayers. Featuring three or four moieties of different philicities, they reach beyond ordinary amphiphilicity and open up avenues to new functions and interaction concepts. Here, we have incorporated a series of X-shaped bolapolyphiles into DPPC (1,2-dipalmitoyl-sn-glycero-3-phosphocholine bilayers of giant unilamellar vesicles. The bolapolyphiles consist of a rod-like oligo(phenylene ethynylene (OPE core, hydrophilic glycerol-based headgroups with or without oligo(ethylene oxide expansions at both ends and two lateral alkyl chains attached near the center of the OPE core. In the absence of DPPC and water, the compounds showed thermotropic liquid-crystalline behavior with a transition between polyphilic and amphiphilic assembly (see part 1 in this issue. In DPPC membranes, various trends in the domain morphologies were observed upon structure variations, which entailed branched alkyl chains of various sizes, alkyl chain semiperfluorination and size expansion of the headgroups. Observed effects on domain morphology are interpreted in the context of the bulk behavior (part 1 and of a model that was previously developed based on spectroscopic and physicochemical data.

Core design of super LWR with double tube water rods

International Nuclear Information System (INIS)

Wu, Jianhui; Oka, Yoshiaki

2014-01-01

Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

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

SNEAK assembly 12B - Construction and experimental program

International Nuclear Information System (INIS)

Helm, F.

1982-10-01

The assembly SNEAK 12A is the second SNEAK assembly in which the neutron physics of disturbed material configurations, as they might occur in a postulated fast reactor accident, is investigated. The core is characterized by a central zone with fuel rods containing PuO 2 UO 2 mixed oxide pellets. This zone is cylindrically surrounded by a buffer zone for spectral adaptation, a driver zone sufficiently reactive to maintain criticality and a radial depleted uranium blanket. The purpose of the experiments is to test calculational methods of neutron physics which are used in the accident analysis of fast breeder reactors. The report describes the construction of the core and its assemblies and their content and outlines the foreseen experimental program. It includes measurements in the clean critical core and investigations of vertical and horizontal fuel relocations and steel blockages

Undermoderated spectrum MOX core study. Pressurized water-type breeder

International Nuclear Information System (INIS)

Tochihara, Hiroshi; Komano, Yasuo

1998-01-01

The purpose of this development is the advance of the PWR core. The conversion ratio (the breeding ratio) are examined. In the case of heavy water as the coolant, the breeding ratio can be achieved about 1.1 with using a hexagonal lattice and space about 1 mm assembly fuel. In the case of light water coolant, the breeding ratio becomes about 1.0, using a hexagonal lattice and fuel space about 0.5 mm fuel assembly. Here, it reports on the situation of the examination such as the nucleus design of core, the design of fuel assembly, the heat hydraulics design of the core, the structure design and so on. (author)

Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

Energy Technology Data Exchange (ETDEWEB)

Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

2015-09-15

Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

Impairment of interferon regulatory factor-3 activation by hepatitis C virus core protein basic amino acid region 1.

Science.gov (United States)

Inoue, Kazuaki; Tsukiyama-Kohara, Kyoko; Matsuda, Chiho; Yoneyama, Mitsutoshi; Fujita, Takashi; Kuge, Shusuke; Yoshiba, Makoto; Kohara, Michinori

2012-11-30

Interferon regulatory factor-3 (IRF-3), a key transcriptional factor in the type I interferon system, is frequently impaired by hepatitis C virus (HCV), in order to establish persistent infection. However, the exact mechanism by which the virus establishes persistent infection has not been fully understood yet. The present study aimed to investigate the effects of various HCV proteins on IRF-3 activation, and elucidate the underlying mechanisms. To achieve this, full-length HCV and HCV subgenomic constructs corresponding to structural and each of the nonstructural proteins were transiently transfected into HepG2 cells. IFN-β induction, plaque formation, and IRF-3 dimerization were elicited by Newcastle disease virus (NDV) infection. The expressions of IRF-3 homodimer and its monomer, Ser386-phosphorylated IRF-3, and HCV core protein were detected by immunofluorescence and western blotting. IFN-β mRNA expression was quantified by real-time PCR (RT-PCR), and IRF-3 activity was measured by the levels of IRF-3 dimerization and phosphorylation, induced by NDV infection or polyriboinosinic:polyribocytidylic acid [poly(I:C)]. Switching of the expression of the complete HCV genome as well as the core proteins, E1, E2, and NS2, suppressed IFN-β mRNA levels and IRF-3 dimerization, induced by NDV infection. Our study revealed a crucial region of the HCV core protein, basic amino acid region 1 (BR1), to inhibit IRF-3 dimerization as well as its phosphorylation induced by NDV infection and poly (I:C), thus interfering with IRF-3 activation. Therefore, our study suggests that rescue of the IRF-3 pathway impairment may be an effective treatment for HCV infection. Copyright © 2012 Elsevier Inc. All rights reserved.

Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

International Nuclear Information System (INIS)

Choi, Hangbok

2014-01-01

The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

New insights into micro/nanoscale combined probes (nanoAuger, μXPS) to characterize Ag/Au@SiO2 core-shell assemblies

Science.gov (United States)

Ledeuil, J. B.; Uhart, A.; Soulé, S.; Allouche, J.; Dupin, J. C.; Martinez, H.

2014-09-01

This work has examined the elemental distribution and local morphology at the nanoscale of core@shell Ag/Au@SiO2 particles. The characterization of such complex metal/insulator materials becomes more efficient when using an initial cross-section method of preparation of the core@shell nanoparticles (ion milling cross polisher). The originality of this route of preparation allows one to obtain undamaged, well-defined and planar layers of cross-cut nano-objects. Once combined with high-resolution techniques of characterization (XPS, Auger and SEM), the process appears as a powerful way to minimize charging effects and enhance the outcoming electron signal (potentially affected by the topography of the material) during analysis. SEM experiments have unambiguously revealed the hollow-morphology of the metal core, while Auger spectroscopy observations showed chemical heterogeneity within the particles (as silver and gold are randomly found in the core ring). To our knowledge, this is the first time that Auger nano probe spectroscopy has been used and successfully optimized for the study of some complex metal/inorganic interfaces at such a high degree of resolution (~12 nm). Complementarily, XPS Au 4f and Ag 3d peaks were finally detected attesting the possibility of access to the whole chemistry of such nanostructured assemblies.This work has examined the elemental distribution and local morphology at the nanoscale of core@shell Ag/Au@SiO2 particles. The characterization of such complex metal/insulator materials becomes more efficient when using an initial cross-section method of preparation of the core@shell nanoparticles (ion milling cross polisher). The originality of this route of preparation allows one to obtain undamaged, well-defined and planar layers of cross-cut nano-objects. Once combined with high-resolution techniques of characterization (XPS, Auger and SEM), the process appears as a powerful way to minimize charging effects and enhance the outcoming

Operation and Licensing of Mixed Cores in Water Cooled Reactors

International Nuclear Information System (INIS)

2013-11-01

Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

Mechanical design of core components for a high performance light water reactor with a three pass core

International Nuclear Information System (INIS)

Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

2007-01-01

Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

2000-03-01

At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

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)

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

International Nuclear Information System (INIS)

Stone, I.Z.

1992-01-01

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

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

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)

Fast Flux Test Facility core system

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

1990-11-01

A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

Synthesis and Plasmonic Understanding of Core/Satellite and Core Shell Nanostructures

Science.gov (United States)

Ruan, Qifeng

Localized surface plasmon resonance, which stems from the collective oscillations of conduction-band electrons, endows Au nanocrystals with unique optical properties. Au nanocrystals possess extremely large scattering/absorption cross-sections and enhanced local electromagnetic field, both of which are synthetically tunable. Moreover, when Au nanocrystals are closely placed or hybridized with semiconductors, the coupling and interaction between the individual components bring about more fascinating phenomena and promising applications, including plasmon-enhanced spectroscopies, solar energy harvesting, and cancer therapy. The continuous development in the field of plasmonics calls for further advancements in the preparation of high-quality plasmonic nanocrystals, the facile construction of hybrid plasmonic nanostructures with desired functionalities, as well as deeper understanding and efficient utilization of the interaction between plasmonic nanocrystals and semiconductor components. In this thesis, I developed a seed-mediated growth method for producing size-controlled Au nanospheres with high monodispersity and assembled Au nanospheres of different sizes into core/satellite nanostructures for enhancing Raman signals. For investigating the interactions between Au nanocrystals and semiconductors, I first prepared (Au core) (TiO2 shell) nanostructures, and then studied their synthetically controlled plasmonic properties and light-harvesting applications. Au nanocrystals with spherical shapes are desirable in plasmon-coupled systems owing to their high geometrical symmetry, which facilitates the analysis of electrodynamic responses in a classical electromagnetic framework and the investigation of quantum tunneling and nonlocal effects. I prepared remarkably uniform Au nanospheres with diameters ranging from 20 nm to 220 nm using a simple seed-mediated growth method associated with mild oxidation. Core/satellite nanostructures were assembled out of differently sized

Reverse depletion method for PWR core reload design

International Nuclear Information System (INIS)

Downar, T.J.; Kim, Y.J.

1985-01-01

Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

Core-shell architectures as nano-size transporters

International Nuclear Information System (INIS)

Adeli, M.; Zarnegar, Z.; Kabiri, R.; Salimi, F.; Dadkah, A.

2006-01-01

Core-shell architectures containing poly (ethylene imine) (PEI) as a core and poly (lactide) (PLA) as arms were prepared. PEI was used as macro initiator for ring opening polymerization of lactide. PEI-PLA core-shell architectures were able to encapsulate guest molecules. Size of the core-shell architectures was between 10- 100 nm, hence they can be considered as nano carriers to transport the guest molecules. Transport capacity of nano carriers depends on their nano-environments and type of self-assembly in solvent. In solid state nano carriers self-assemble as long structures with nano-size diameter or they form network structures. Aggregations type depends on the concentration of nano carriers in solution. Effect of the shell thickness and aggregation type on the release rate are also investigated

LMFR core thermohydraulics: Status and prospects

International Nuclear Information System (INIS)

2000-06-01

One of the fundamental steps for a successful reactor core thermohydraulic design is the capability to predict, reliably and accurately, the temperature distribution in the core assemblies. A detailed knowledge of the assembly and fuel pin thermohydraulic behaviour in the steady state and transient conditions is an indispensable prerequisite to safe and stable operation of the reactor. Considerable experimental and theoretical studies on various aspects of LMFR core thermohydraulics are necessary to acquire such knowledge. During the last decade, there have been substantial advances in fast reactor core thermohydraulic design and operation in several countries with fast reactor programmes (notably in France, the Russian Federation, Japan, the United Kingdom, Germany and the United States of America). Chief among these has been the demonstration of reliable operation of reactor cores at a high burnup. During the last years, some additional countries such as China, India and the Republic of Korea have launched new fast reactor programmes. International exchange of information and experience on LMFR development including core thermohydraulic design is becoming of increasing importance to these countries. It is with this focus that the IAEA convened the Technical Committee on 'Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR's with Traditional and Burner Cores'. This meeting, attended by participants from seven countries, brought together a group of international experts to review and discuss the thermohydraulic advances and design approaches providing a reliable, safe and robust reactor core, as well as to exchange the experience accumulated in different countries of using the codes for thermohydraulic calculations and to discuss the issues requiring further research and development. A total of thirty technical papers presented covered theoretical and computational issues as well as experiments under

Design improvement for fretting-wear reduction of HANARO fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Design improvement for fretting-wear reduction of HANARO fuel assembly

International Nuclear Information System (INIS)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R.

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Evaluation of in-core measurements by means of principal components method

International Nuclear Information System (INIS)

Makai, M.; Temesvari, E.

1996-01-01

Surveillance of a nuclear reactor core comprehends determination of assemblies' three-dimensional (3D) power distribution. Derived from other assemblies' measured values, power of non-measured assembly is calculated for every assembly with the help of principal components method (PCM) which is also presented. The measured values are interpolated for different geometrical coverings of the WWER-440 core. Different procedures have been elaborated and investigated, among them the most successful methods are discussed. Each method offers self consistent means to determine numerical errors of the interpolated values. (author). 13 refs, 7 figs, 2 tabs

Safety considerations of new critical assembly for the Research Reactor Institute, Kyoto University

International Nuclear Information System (INIS)

Umeda, Iwao; Matsuoka, Naomi; Harada, Yoshihiko; Miyamoto, Keiji; Kanazawa, Takashi

1975-01-01

The new critical assembly type of nuclear reactor having three cores for the first time in the world was completed successfully at the Research Reactor Institute of Kyoto University in autumn of 1974. It is called KUCA (Kyoto University Critical Assembly). Safety of the critical assembly was considered sufficiently in consequence of discussions between the researchers of the institute and the design group of our company, and then many bright ideas were created through the discussions. This paper is described the new safety design of main equipments - oil pressure type center core drive mechanism, removable water overflow mechanism, core division mechanism, control rod drive mechansim, protection instrumentation system and interlock key system - for the critical assembly. (author)

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)

Some characteristics of two heterogeneous cores and their experimental confirmation

International Nuclear Information System (INIS)

Giese, H.; Henneges, G.; Pilate, S.

1979-01-01

Heterogeneous core geometries have been investigated at the zero power facility ZEBRA with the objective of developing an improved understanding of the basic physics parameters of future non-conventional fast breeder reactors. Three assemblies were investigated, comprising similar fissile volumes and numbers of in-core fertile elements. First, BZC, a Salt-and-Pepper Core with fertile elements arranged in groups of sizes ranging from about one to eight power reactor subassemblies. Second, BZD, a Single Annular Core with the fertile elements collected to form a large central island of about 95 cm diameter, equivalent to about 36 subassemblies. Third, BZD/1A, a modified version of assembly BZD with a central island diameter of about 67 cm and an additional thin breeder ring. All assemblies used 24% Pu/Pu + U enriched fuel, with a core height of about 90 cm and a core diameter of about 200 cm. Extensive sodium voiding, flux-tilt and reactivity-interaction measurements have been carried out. The results indicate that relative to homogeneous designs a Single Annular arrangement yields a more marked improvement in sodium voiding performance than a Salt-and-Pepper arrangement, at the expense, however, of an increased sensitivity of the flux distribution to local reactivity perturbations. First analysis-level calculations on sodium-void yield C/E values with a significantly wider dispersion than in previous homogeneous assemblies. (author)

Discussion on the re-irradiated fuel assembly with damaged guide vanes

International Nuclear Information System (INIS)

Li Ligang

2013-01-01

In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

Core Seismic Tests for a Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Lee, J. H

2007-01-15

This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

Monitoring device for the power distribution within a nuclear reactor core

International Nuclear Information System (INIS)

Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

1986-01-01

Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

Top-nozzle mounted replacement guide pin assemblies

International Nuclear Information System (INIS)

Gilmore, C.B.; Andrews, W.H.

1993-01-01

A replacement guide pin assembly is provided for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core. The guide pin assembly includes a guide pin body having a radially expandable base insertable within a hole in the top nozzle, a ferrule insertable within the guide pin base and capable of imparting a radially and outwardly directed force on the expandable base to expand it within the hole of the top nozzle and thereby secure the guide pin body to the top nozzle in response to a predetermined displacement of the ferrule relative to the guide pin body along its longitudinal axis, and a lock screw interfitted with the ferrule and threaded into the guide pin body so as to produce the predetermined displacement of the ferrule. (author)

Lateral restraint assembly in a nuclear reactor

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, W.

1977-01-01

A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

Compaction of a Bacterial Group I Ribozyme Coincides with the Assembly of Core Helices

International Nuclear Information System (INIS)

Perez-Salas, U.; Rangan, P.; Krueger, S.; Briber, R.; Thirumalai, D.; Woodson, S.

2004-01-01

Counterions are critical to the self-assembly of RNA tertiary structure because they neutralize the large electrostatic forces which oppose the folding process. Changes in the size and shape of the Azoarcus group I ribozyme as a function of Mg 2+ and Na + concentration were followed by small angle neutron scattering. In low salt buffer, the RNA was expanded, with an average radius of gyration (R g ) of 53 ± 1 Angstroms. A highly cooperative transition to a compact form (R g = 31.5 ± 0.5 Angstroms) was observed between 1.6 and 1.7 mM MgCl 2 . The collapse transition, which is unusually sharp in Mg 2+ , has the characteristics of a first-order phase transition. Partial digestion with ribonuclease T1 under identical conditions showed that this transition correlated with the assembly of double helices in the ribozyme core. Fivefold higher Mg 2+ concentrations were required for self-splicing, indicating that compaction occurs before native tertiary interactions are fully stabilized. No further decrease in Rg was observed between 1.7 and 20 mM MgCl 2 , indicating that the intermediates have the same dimensions as the native ribozyme, within the uncertainty of the data (± 1 Angstroms). A more gradual transition to a final R g of approximately 33.5 Angstroms was observed between 0.45 and 2 M NaCl. This confirms the expectation that monovalent ions not only are less efficient in charge neutralization but also contract the RNA less efficiently than multivalent ions

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)

Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

International Nuclear Information System (INIS)

Hermansky, J.; Petenyi, V.; Zavodsky, M.

2010-01-01

The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

Internal core tightener

International Nuclear Information System (INIS)

Brynsvold, G.V.; Snyder, H.J. Jr.

1976-01-01

An internal core tightener is disclosed which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a ''fixed'' outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change. 5 claims, 12 drawing figures

Whole core transport calculation for the VHTR hexagonal core

International Nuclear Information System (INIS)

Cho, J. Y.; Kim, K. S.; Lee, C. C.; Joo, H. G.

2007-01-01

Recently, the DeCART code which performs the whole core calculation by coupling the radial MOC transport kernel with the axial nodal kernel has equipped a kernel to deal with the hexagonal geometry and applied to the VHTR hexagonal core to examine the accuracy and the computational efficiency of the implemented kernel. The implementation includes a modular ray tracing module based on the hexagonal assembly and a multi-group CMFD module to perform an efficient transport calculation. The requirements for the modular ray are: (1) the assembly based path linking and (2) the complete reflection capabilities. The first requirement is met by adjusting the azimuthal angle and the ray spacing for the modular ray to construct a core ray by the path linking. The second requirement is met by expanding the constructed azimuthal angle in the range of [0,30 degree] to the remained range to reflect completely at the core boundaries. The considered reflecting surface angles for the complete reflection are 30n's (n=1,2,1,12). The CMFD module performs the equivalent diffusion calculation to the radial MOC transport calculation based on the homogenized structure units. The structure units include the hexagonal pin cells and gap cells appearing at the assembly boundary. Therefore, the CMFD module is programmed to deal with the unstructured cells such as the gap cells. The CMFD equation consists of the two parts of (1) the conventional FDM and (2) the current corrective parts. Since the second part of the CMFD equation guarantees the reproducibility of the radial MOC transport solutions for the cell averaged reaction rate and the net current at the cell surfaces, how to build the first part of the CMFD equation is not important. Therefore, the first part of the CMFD equation is roughly built by using the normal distance from the gravity center to the surface. The VHTR core uses helium as a coolant which is realized as a void hole in a neutronics calculation. This void hole which

The verification of PWR-fuel code for PWR in-core fuel management

International Nuclear Information System (INIS)

Surian Pinem; Tagor M Sembiring; Tukiran

2015-01-01

In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

Thermal behavior of molten corium during TMI-2 core relocation event

International Nuclear Information System (INIS)

Anderson, J.L.; Sienicki, J.J.

1988-01-01

During the TMI-2 accident, a pool of molten corium formed in the central region of the core and was contained by solidified crusts. Failure of the crust surrounding the molten material, at approximately 224 min, resulted in a relocation of an estimated 20-25 tons of molten corium through peripheral fuel assemblies in the east side of the vessel, as well as through the core barrel assembly (CBA) at the periphery of the core. This paper presents the results of an analyses carried out to investigate the thermal interactions of molten corium with the CBA structures during the relocation event. The principal objectives of the analyses are: (a) to assess the potential for relocation to take place through the CBA versus the flow of molten core material directly downward through the core via the fuel assemblies; and (b) to understand the distribution of prior molten corium observed during vessel defueling examinations. 5 refs., 1 fig

Synthesis of core-fluorescent four-armed star and dicyclic 8-shaped poly(THF)s by electrostatic self-assembly and covalent fixation (ESA–CF) protocol

KAUST Repository

Fujiwara, Susumu; Yamamoto, Takuya; Tezuka, Yasuyuki; Habuchi, Satoshi

2013-01-01

A pair of four-armed star and dicyclic 8-shaped poly(tetrahydrofuran)s, poly(THF)s, possessing a perylene diimide group at the core position (Ia and Ib, respectively) were synthesized by means of an electrostatic self-assembly and covalent fixation (ESA–CF) protocol. Mono- and bifunctional poly(THF)s having N-phenylpiperidinium salt end groups accompanying a perylene diimide tetracarboxylate as a counteranion were prepared by the ion-exchange reaction, and the subsequent covalent conversion by reflux in toluene afforded the corresponding core-fluorescent four-armed star and dicyclic 8-shaped poly(THF)s, (Ia and Ib, respectively) for the use of single-molecule fluorescence microscopy measurements.

Synthesis of core-fluorescent four-armed star and dicyclic 8-shaped poly(THF)s by electrostatic self-assembly and covalent fixation (ESA–CF) protocol

KAUST Repository

Fujiwara, Susumu

2013-12-07

A pair of four-armed star and dicyclic 8-shaped poly(tetrahydrofuran)s, poly(THF)s, possessing a perylene diimide group at the core position (Ia and Ib, respectively) were synthesized by means of an electrostatic self-assembly and covalent fixation (ESA–CF) protocol. Mono- and bifunctional poly(THF)s having N-phenylpiperidinium salt end groups accompanying a perylene diimide tetracarboxylate as a counteranion were prepared by the ion-exchange reaction, and the subsequent covalent conversion by reflux in toluene afforded the corresponding core-fluorescent four-armed star and dicyclic 8-shaped poly(THF)s, (Ia and Ib, respectively) for the use of single-molecule fluorescence microscopy measurements.

Electrostatics and the assembly of an RNA virus

NARCIS (Netherlands)

Schoot, van der P.P.A.M.; Bruinsma, R.

2005-01-01

Electrostatic interactions play a central role in the assembly of single-stranded RNA viruses. Under physiological conditions of salinity and acidity, virus capsid assembly requires the presence of genomic material that is oppositely charged to the core proteins. In this paper we apply basic polymer

Colloidal Gold--Collagen Protein Core--Shell Nanoconjugate: One-Step Biomimetic Synthesis, Layer-by-Layer Assembled Film, and Controlled Cell Growth.

Science.gov (United States)

Xing, Ruirui; Jiao, Tifeng; Yan, Linyin; Ma, Guanghui; Liu, Lei; Dai, Luru; Li, Junbai; Möhwald, Helmuth; Yan, Xuehai

2015-11-11

The biogenic synthesis of biomolecule-gold nanoconjugates is of key importance for a broad range of biomedical applications. In this work, a one-step, green, and condition-gentle strategy is presented to synthesize stable colloidal gold-collagen core-shell nanoconjugates in an aqueous solution at room temperature, without use of any reducing agents and stabilizing agents. It is discovered that electrostatic binding between gold ions and collagen proteins and concomitant in situ reduction by hydroxyproline residues are critically responsible for the formation of the core-shell nanoconjugates. The film formed by layer-by-layer assembly of such colloidal gold-collagen nanoconjugates can notably improve the mechanical properties and promote cell adhesion, growth, and differentiation. Thus, the colloidal gold-collagen nanoconjugates synthesized by such a straightforward and clean manner, analogous to a biomineralization pathway, provide new alternatives for developing biologically based hybrid biomaterials toward a range of therapeutic and diagnostic applications.

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

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Core barrel inner tube lifter

Energy Technology Data Exchange (ETDEWEB)

Jeffers, J P

1968-07-16

A core drill with means for selectively lifting a core barrel inner tube consists of a lifting means connected to the core barrel inner tube assembly. It has a closable passage to permit drilling fluid normally to pass through it. The lifting means has a normally downward facing surface and a means to direct drilling fluid pressure against that surface so that on closure of the passage to fluid flow, the pressure of the drilling fluid is caused to act selectively on it. This causes the lifting means to rise and lift the core barrel. (7 claims)

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

Au nanorice assemble electrolytically into mesostars.

Science.gov (United States)

Bardhan, Rizia; Neumann, Oara; Mirin, Nikolay; Wang, Hui; Halas, Naomi J

2009-02-24

Star-shaped mesotructures are formed when an aqueous suspension of Au nanorice particles, which consist of prolate hematite cores and a thin Au shell, is subjected to an electric current. The nanorice particles assemble to form hyperbranched micrometer-scale mesostars. To our knowledge, this is the first reported observation of nanoparticle assembly into larger ordered structures under the influence of an electrochemical process (H(2)O electrolysis). The assembly is accompanied by significant modifications in the morphology, dimensions, chemical composition, crystallographic structure, and optical properties of the constituent nanoparticles.

Nuclear reactor core servicing apparatus

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

Device for handling fuel assemblies within a nuclear reactor core

International Nuclear Information System (INIS)

Dupuy, G.

1975-01-01

A device is described which comprises two arms having synchronized movements, the tubular portions of the two arms being rigidly fixed to each other by means of a sliding connection and capable of being endowed by means of a differential drive system with movements for producing the same effort but applied in the opposite direction in order that the lateral reaction on the grab of the first arm on a fuel assembly should be equal and opposite to the lateral reaction of the cylindrical guide tube of the second arm on the same fuel assembly

Neutron transport assembly calculation with non-zero net current boundary condition

International Nuclear Information System (INIS)

Jo, Chang Keun

1993-02-01

Fuel assembly calculation for the homogenized group constants is one of the most important parts in the reactor core analysis. The homogenized group constants of one a quarter assembly are usually generated for the nodal calculation of the reactor core. In the current nodal calculation, one or a quarter of the fuel assembly corresponds to a unit node. The homogenized group constant calculation for a fuel assembly proceeds through cell spectrum calculations, group condensation and cell homogenization calculations, two dimensional fuel assembly calculation, and then depletion calculations of fuel rods. To obtain the assembly wise homogenized group constants, the two dimensional transport calculation is usually performed. Most codes for the assembly wise homogenized group constants employ a zero net current boundary condition. CASMO-3 is such a code that is in wide use. The zero net current boundary condition is plausible and valid in an infinite reactor composed of the same kind of assemblies. However, the reactor is finite and the core is constructed by different kinds of assemblies. Hence, the assumption of the zero net current boundary condition is not valid in the actual reactor. The objective of this study is to develop a homogenization methodology that can treat any actual boundary condition, i.e. non-zero net current boundary condition. In order to treat the non-zero net current boundary condition, we modify CASMO-3. For the two-dimensional treatment in CASMO-3, a multigroup integral transport routine based on the method of transmission probability is used. The code performs assembly calculation with zero net current boundary condition. CASMO-3 is modified to consider the inhomogeneous source at the assembly boundary surface due to the non-zero net current. The modified version of CASMO-3 is called CASMO-3M. CASMO-3M is applied to several benchmark problems. In order to obtain the inhomogeneous source, the global calculation is performed. The local calculation

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

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)

Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

International Nuclear Information System (INIS)

Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

1985-01-01

The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

2016-12-15

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Fluid moderator control system fuel assembly seal connector

International Nuclear Information System (INIS)

Veronesi, L.; Tower, S.N.; Klassen, W.E.

1987-01-01

A nuclear reactor is described comprising fuel assemblies having one or more flow channels therethrough, a core support plate having one or more flow channels therethrough, and seal connectors for sealingly connecting the one or more flow channels in the core support plate with the one or more flow channels in the fuel assemblies. The seal connectors each comprises a first portion and second portion each comprising an elongated member having a flow channel therethrough and being in substantial axial alignment with each other and being separated by a space therebetween, means for sealingly connecting the first portion o one or the one or more flow channels in the fuel assemblies, means for sealingly connecting the second portion to the first portion and for allowing relative motion between the portions, means for limiting the relative motion of the first and second portion in directions toward and away from each other, means for reconnectingly connecting and resealingly sealing the second portion to one of the one or more flow channels in the core support plate. It comprises a slip fit connection whereby the remote end of the second portion fits within whereby the remote end of the second portion fits within an opening in the core support plate which is in flow communication with the one or more flow channels in the core support plate and further comprises a ball and cones seal in series with axially spaced ring seals

Analysis of hepatitis C virus core/NS5A protein co-localization using novel cell culture systems expressing core-NS2 and NS5A of genotypes 1-7

DEFF Research Database (Denmark)

Galli, Andrea; Scheel, Troels K H; Prentoe, Jannick C

2013-01-01

Hepatitis C virus (HCV) is an important human pathogen infecting hepatocytes. With the advent of infectious cell culture systems, the HCV particle assembly and release processes are finally being uncovered. The HCV core and NS5A proteins co-localize on cytoplasmic lipid droplets (c......LDs) or on the endoplasmic reticulum (ER) at different stages of particle assembly. Current knowledge on assembly and release is primarily based on studies in genotype 2a cell culture systems; however, given the high genetic heterogeneity of HCV, variations might exist among genotypes. Here, we developed novel HCV strain...... JFH1-based recombinants expressing core-NS2 and NS5A from genotypes 1-7, and analysed core and NS5A co-localization in infected cells. Huh7.5 cells were transfected with RNA of core-NS2/NS5A recombinants and putative adaptive mutations were analysed by reverse genetics. Adapted core-NS2/NS5A...

Less symptomatic, but equally impaired: Clinical impairment in restricting versus binge-eating/purging subtype of anorexia nervosa.

Science.gov (United States)

Reas, Deborah Lynn; Rø, Øyvind

2018-01-01

This study investigated subtype differences in eating disorder-specific impairment in a treatment-seeking sample of individuals with anorexia nervosa (AN). The Clinical Impairment Assessment (CIA) and the Eating Disorder Examination-Questionnaire (EDE-Q) were administered to 142 patients. Of these, 54.9% were classified as restricting type (AN-R) and 45.1% were classified as binge-eating/purging type (AN-B/P) based on an average weekly occurrence of binge eating and/or purging episodes (≥4 episodes/28days). Individuals with AN-B/P exhibited higher levels of core ED psychopathology (dietary restraint, eating concern, shape/weight concerns) in addition to the expected higher frequency of binge/purge episodes. No significant differences existed between AN subtypes in the severity of ED-related impairment. Weight/shape concerns and binge eating frequency significantly predicted level of impairment. Differential associations were observed between the type of ED pathology that significantly contributed to impairment according to AN subtype. Although those with AN-B/P displayed higher levels of core attitudinal and behavioral ED pathology than AN-R, no significant differences in ED-specific impairment were found between AN subtypes. Eating disorder-related impairment in AN was not related to the severity of underweight or purging behaviors, but was uniquely and positively associated with weight/shape concerns and binge eating frequency. Copyright © 2018 Elsevier Ltd. All rights reserved.

A reverse depletion method for pressurized water reactor core reload design

International Nuclear Information System (INIS)

Downar, T.J.; Kin, Y.J.

1986-01-01

Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. The large numbers of burnable poison pins used to control the power peaking at the in-board fresh fuel positions have introduced an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of burnable poison pins in the fresh fuel. A new method for performing core design more suitable for low-leakage fuel management is reported. A procedure was developed that uses the wellknown ''Haling depletion'' to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and burnable poison distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provided for the maximum utility to PWR core reload design

Study on Reactor Physics Characteristic of the PWR Core Using UO2

International Nuclear Information System (INIS)

Tukiran Surbakti

2009-01-01

Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Finding Nemo's Genes: A chromosome-scale reference assembly of the genome of the orange clownfish Amphiprion percula

KAUST Repository

Lehmann, Robert; Lightfoot, Damien J; Schunter, Celia Marei; Michell, Craig T; Ohyanagi, Hajime; Mineta, Katsuhiko; Foret, Sylvain; Berumen, Michael L.; Miller, David J; Aranda, Manuel; Gojobori, Takashi; Munday, Philip L; Ravasi, Timothy

2018-01-01

The iconic orange clownfish, Amphiprion percula, is a model organism for studying the ecology and evolution of reef fishes, including patterns of population connectivity, sex change, social organization, habitat selection and adaptation to climate change. Notably, the orange clownfish is the only reef fish for which a complete larval dispersal kernel has been established and was the first fish species for which it was demonstrated that anti-predator responses of reef fishes could be impaired by ocean acidification. Despite its importance, molecular resources for this species remain scarce and until now it lacked a reference genome assembly. Here we present a de novo chromosome-scale assembly of the genome of the orange clownfish Amphiprion percula. We utilized single-molecule real-time sequencing technology from Pacific Biosciences to produce an initial polished assembly comprised of 1,414 contigs, with a contig N50 length of 1.86 Mb. Using Hi-C based chromatin contact maps, 98% of the genome assembly were placed into 24 chromosomes, resulting in a final assembly of 908.8 Mb in length with contig and scaffold N50s of 3.12 and 38.4 Mb, respectively. This makes it one of the most contiguous and complete fish genome assemblies currently available. The genome was annotated with 26,597 protein coding genes and contains 96% of the core set of conserved actinopterygian orthologs. The availability of this reference genome assembly as a community resource will further strengthen the role of the orange clownfish as a model species for research on the ecology and evolution of reef fishes.

Finding Nemo's Genes: A chromosome-scale reference assembly of the genome of the orange clownfish Amphiprion percula

KAUST Repository

Lehmann, Robert

2018-03-08

The iconic orange clownfish, Amphiprion percula, is a model organism for studying the ecology and evolution of reef fishes, including patterns of population connectivity, sex change, social organization, habitat selection and adaptation to climate change. Notably, the orange clownfish is the only reef fish for which a complete larval dispersal kernel has been established and was the first fish species for which it was demonstrated that anti-predator responses of reef fishes could be impaired by ocean acidification. Despite its importance, molecular resources for this species remain scarce and until now it lacked a reference genome assembly. Here we present a de novo chromosome-scale assembly of the genome of the orange clownfish Amphiprion percula. We utilized single-molecule real-time sequencing technology from Pacific Biosciences to produce an initial polished assembly comprised of 1,414 contigs, with a contig N50 length of 1.86 Mb. Using Hi-C based chromatin contact maps, 98% of the genome assembly were placed into 24 chromosomes, resulting in a final assembly of 908.8 Mb in length with contig and scaffold N50s of 3.12 and 38.4 Mb, respectively. This makes it one of the most contiguous and complete fish genome assemblies currently available. The genome was annotated with 26,597 protein coding genes and contains 96% of the core set of conserved actinopterygian orthologs. The availability of this reference genome assembly as a community resource will further strengthen the role of the orange clownfish as a model species for research on the ecology and evolution of reef fishes.

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Extreme-Scale De Novo Genome Assembly

Energy Technology Data Exchange (ETDEWEB)

Georganas, Evangelos [Intel Corporation, Santa Clara, CA (United States); Hofmeyr, Steven [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Joint Genome Inst.; Egan, Rob [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Computational Research Division; Buluc, Aydin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Joint Genome Inst.; Oliker, Leonid [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Joint Genome Inst.; Rokhsar, Daniel [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Computational Research Division; Yelick, Katherine [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Joint Genome Inst.

2017-09-26

De novo whole genome assembly reconstructs genomic sequence from short, overlapping, and potentially erroneous DNA segments and is one of the most important computations in modern genomics. This work presents HipMER, a high-quality end-to-end de novo assembler designed for extreme scale analysis, via efficient parallelization of the Meraculous code. Genome assembly software has many components, each of which stresses different components of a computer system. This chapter explains the computational challenges involved in each step of the HipMer pipeline, the key distributed data structures, and communication costs in detail. We present performance results of assembling the human genome and the large hexaploid wheat genome on large supercomputers up to tens of thousands of cores.

Study of advanced LWR cores for effective use of plutonium and MOX physics experiments

International Nuclear Information System (INIS)

Yamamoto, T.; Matsu-Ura, H.; Ueji, M.; Ota, H.; Kanagawa, T.; Sakurada, K.; Maruyama, H.

1999-01-01

Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ∼5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ∼6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. (author)

Reaction Rate Distributions and Ratios in FR0 Assemblies 1, 2 and 3

Energy Technology Data Exchange (ETDEWEB)

Andersson, T L

1966-06-15

The spatial distribution of different reaction rates and reaction ratios in Assemblies 1, 2 and 3 of the fast reactor FR0 was measured by fission chamber scans and foil activation technique. Assemblies 1 and 2 had cores of undiluted fuel (uranium metal enriched to 20 % U{sup 235}) while the core of Assembly 3 was diluted with about 30 vol. % graphite. All the systems had a thick copper reflector, The experimental results were compared with calculated values obtained from DSN and TDC multigroup spectra and group cross-section sets for the reactions. Good agreement between experiment and calculations is generally obtained in the core region but in the reflector the neutron spectrum is calculated too hard.

Power peak in vicinity of WWER-440 control assembly

International Nuclear Information System (INIS)

Mikus, J.

2002-01-01

This paper presents information concerning the WWER-440 local power peaking problem induced by a control assembly and corresponding investigation possibilities on the light-water zero-power reactor LR-O at the Nuclear Research Institute Rez plc. Brief description of the disposable CA model, experimental arrangement and conditions on the LR-O reactor, preparation of the relevant measurements in the WWER-440 type cores with CA model, as well as some preliminary results of the fission density distribution obtained in a core without boron and with fuel assemblies having profiled enrichment are mentioned too (Author)

In-core monitoring detectors

International Nuclear Information System (INIS)

Mitelman, M.G.

2001-01-01

The main task of in-core monitoring consists in securing observability of the reactor installation in all possible operation modes (normal, transient, accident and post-accident). Operation safety at acceptable cost can be achieved by optimized measurement errors. The range of sensors applied as in-core detectors for operative measurements in the industry is very limited in number. Among them might be cited self powered neutron detectors (SPND) and thermocouples. Sensors are incorporated in the in-core detectors assemblies (SVRD). The presentation makes an effort to touch upon the problems of assuring and increasing quality of in-core on-line measurements. So we do not consider systems using movable detectors, as the latter do not assure on-line measurements. (Authors)

Seismic behaviour of PWR fuel assemblies model and its validation

International Nuclear Information System (INIS)

Queval, J.C.; Gantenbein, F.; Brochard, D.; Benjedidia, A.

1991-01-01

The validity of the models simulating the seismic behaviour of PWR cores can only be exactly demonstrated by seismic testing on groups of fuel assemblies. Shake table seismic tests of rows of assembly mock-ups, conducted by the CEA in conjunction with FRAMATOME, are presented in reference /1/. This paper addresses the initial comparisons between model and test results for a row of five assemblies in air. Two models are used: a model with a single beam per assembly, used regularly in accident analyses, and described in reference /2/, and a more refined 2-beam per assembly model, geared mainly towards interpretation of test results. The 2-beam model is discussed first, together with parametric studies used to characterize it, and the study of the assembly row for a period limited to 2 seconds and for different excitation levels. For the 1-beam model assembly used in applications, the row is studied over the total test time, i.e twenty seconds, which covers the average duration of the core seismic behaviour studies, and for a peak exciting acceleration value at 0.4 g, which corresponds to the SSE level of the reference spectrum

Application of radcal gamma thermometer assemblies for core coolant monitoring in ASEA ATOM reactors with particular reference to the Barsebaeck plants

International Nuclear Information System (INIS)

Romslo, K.

1982-02-01

In this study reference designs for instrument assemblies containing RGT rods to monitor the core coolant conditions in the Barsebaeck reactors have been worked out. Four such strings would be required to satisfy the Reg. Guide 1.97 reqiurements. The signal transmission to the control room and the presentation of information to the operators have been addressed. Downcomer water level measurement is considered important in order to get an early warning about leakages. Possible ways of diversifying the existing measurement method using RGTs are mentioned, and the design of a downcomer RGT rod has been suggested. To fully comply with Reg. Guide 1.97, water level measurements above core would be required. In a conceptual way it has been shown how an RGT rod could be extended up into this region, if so required. The possibility of making an ideal core coolant monitoring system by replacing one of the structural rods (water rods) in the fuel bundle by an RGT rod is pointed out. There are foreseen, however, several practical obstacles in pursuing the idea. The present state of RGT development and further work required to get the intrument licensed as a coolant monitoring device, has been defined. (Author)

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2009-01-01

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2008-01-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

On numerical simulation of fuel assembly bow in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Horváth, Ákos, E-mail: akoshorvath@t-online.hu [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Budapest University of Technology and Economics, Department of Aircraft and Ships, Stoczek Street 6, Building J, H-1111 Budapest (Hungary); Dressel, Bernd [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

2013-12-15

Highlights: • Simulation of fuel assembly bow by coupled CFD and finite element method. • Comparison of calculated and experimentally measured bow shapes. • Investigation of boundary condition effect on bow pattern of a fuel assembly row. • Highlighting importance of consideration of fluid–structure interaction. • Assessment of flow redistribution within the fuel assembly row model. - Abstract: Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of

Advanced BWR core component designs and the implications for SFD analysis

International Nuclear Information System (INIS)

Ott, L.J.

1997-01-01

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

Intra-assembly flow redistribution in LMFBRs: a simple computational approach

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.G.

1983-01-01

The liquid metal fast breeder reactor (LMFBR) core consists of fuel, blanket, control, and shielding assemblies packed in a hexagonal configuration. Radial blanket assemblies occupy peripheral locations in the reactor core and are characterized by steep power gradients, while inner blanket assemblies are located within the fuel assembly region and have higher power levels but flatter distributions. It is due to the presence of this radial power gradient that large sodium temperature distributions exist at full power operation. However, at low power, low flow natural convection conditions, a significant flow redistribution takes place leading to considerable radial temperature flattening. The purpose of the present study is to formulate a simple flow-regime dependent model supported by experimental data for prediction of sodium temperature flattening due to buoyancy-induced flow redistribution in LMFBR subassemblies with significant radial power gradient

Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

Ideas to Design an in situ Diamond Drilling Core Splitter within Soft ...

African Journals Online (AJOL)

Michael O. Mensah

2015-12-02

Dec 2, 2015 ... the wireline system of core barrel assembly and the device used in splitting of core ... Keywords: Design, In situ, Diamond drilling, Core splitter, Wireline system .... This is the most complex part of the core barrel and has many.

Advance of core design method for ATR

International Nuclear Information System (INIS)

Maeda, Seiichirou; Ihara, Toshiteru; Iijima, Takashi; Seino, Hideaki; Kobayashi, Tetsurou; Takeuchi, Michio; Sugawara, Satoru; Matsumoto, Mitsuo.

1995-01-01

Core characteristics of ATR demonstration plant has been revised such as increasing the fuel burnup and the channel power, which is achieved by changing the number of fuel rod per fuel assembly from 28 to 36. The research and development concerning the core design method for ATR have been continued. The calculational errors of core analysis code have been evaluated using the operational data of FUGEN and the full scale simulated test results in DCA (Deuterium Critical Assembly) and HTL (Heat Transfer Loop) at O-arai engineering center. It is confirmed that the calculational error of power distribution is smaller than the design value of ATR demonstration plant. Critical heat flux correlation curve for 36 fuel rod cluster has been developed and the probability evaluation method based on its curve, which is more rational to evaluate the fuel dryout, has been adopted. (author)

Plutonium assemblies in reload 1 of the Dodewaard Reactor

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

1977-01-01

Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

Development of UCMS for Analysis of Designed and Measured Core Power Distribution

International Nuclear Information System (INIS)

Moon, Sang Rae; Hong, Sun Kwan; Yang, Sung Tae

2009-01-01

In this study, reactor core loading patterns were determined by calculating and verifying the factors affecting peak power and important core safety variables were reconciled with their design criteria using a newly designed unified core management system. Core loading patterns are designed for quadrant cores under the assumption that the power distribution of the reactor core is the same among symmetric fuel assemblies within the core. Actual core power distributions measured during core operation may differ slightly from their designed data. Reactor engineers monitor these differences between the designed and measured data by performing a surveillance procedure every month according to the technical specification requirements. It is difficult to monitor overall power distribution behavior throughout the assemblies using the current procedure because it requires the reactor engineer to compare the designed data with only the maximum value of the power peaking factor and the relative power density. It is necessary to enhance this procedure to check the primary variables such as core power distribution, because long cycle operation, high burnup, power up-rate, and improved fuel can change the environment in the core. To achieve this goal, a web-based Unified Core Management System (UCMS) was developed. To build the UCMS, a database system was established using reactor design data such as that in the Nuclear Design Report (NDR) and automated core analysis codes for all light water reactor power plants. The UCMS is designed to help reactor engineers to monitor important core variables and core safety margins by comparing the measured core power distribution with designed data for each fuel assembly during the cycle operation in nuclear power plants

A simple and versatile system for the ATP-dependent assembly of chromatin.

Science.gov (United States)

Khuong, Mai T; Fei, Jia; Cruz-Becerra, Grisel; Kadonaga, James T

2017-11-24

Chromatin is the natural form of DNA in the eukaryotic nucleus and is the substrate for diverse biological phenomena. The functional analysis of these processes ideally would be carried out with nucleosomal templates that are assembled with customized core histones, DNA sequences, and chromosomal proteins. Here we report a simple, reliable, and versatile method for the ATP-dependent assembly of evenly spaced nucleosome arrays. This minimal chromatin assembly system comprises the Drosophila nucleoplasmin-like protein (dNLP) histone chaperone, the imitation switch (ISWI) ATP-driven motor protein, core histones, template DNA, and ATP. The dNLP and ISWI components were synthesized in bacteria, and each protein could be purified in a single step by affinity chromatography. We show that the dNLP-ISWI system can be used with different DNA sequences, linear or circular DNA, bulk genomic DNA, recombinant or native Drosophila core histones, native human histones, the linker histone H1, the non-histone chromosomal protein HMGN2, and the core histone variants H3.3 and H2A.V. The dNLP-ISWI system should be accessible to a wide range of researchers and enable the assembly of customized chromatin with specifically desired DNA sequences, core histones, and other chromosomal proteins. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

Nanoporous Network Channels from Self-Assembled Triblock Copolymer Supramolecules

NARCIS (Netherlands)

du Sart, Gerrit Gobius; Vukovic, Ivana; Vukovic, Zorica; Polushkin, Evgeny; Hiekkataipale, Panu; Ruokolainen, Janne; Loos, Katja; ten Brinke, Gerrit

2011-01-01

Supramolecular complexes of a poly(tert-butoxystyrene)-block-polystyrene-block-poly(4-vinylpyridine) triblock copolymers and less than stoichiometric amounts of pentadecylphenol (PDP) are shown to self-assemble into a core-shell gyroid morphology with the core channels formed by the hydrogen-bonded

Benchmark calculations of power distribution within assemblies

International Nuclear Information System (INIS)

Cavarec, C.; Perron, J.F.; Verwaerde, D.; West, J.P.

1994-09-01

The main objective of this Benchmark is to compare different techniques for fine flux prediction based upon coarse mesh diffusion or transport calculations. We proposed 5 ''core'' configurations including different assembly types (17 x 17 pins, ''uranium'', ''absorber'' or ''MOX'' assemblies), with different boundary conditions. The specification required results in terms of reactivity, pin by pin fluxes and production rate distributions. The proposal for these Benchmark calculations was made by J.C. LEFEBVRE, J. MONDOT, J.P. WEST and the specification (with nuclear data, assembly types, core configurations for 2D geometry and results presentation) was distributed to correspondents of the OECD Nuclear Energy Agency. 11 countries and 19 companies answered the exercise proposed by this Benchmark. Heterogeneous calculations and homogeneous calculations were made. Various methods were used to produce the results: diffusion (finite differences, nodal...), transport (P ij , S n , Monte Carlo). This report presents an analysis and intercomparisons of all the results received

Fuel assembly for pressure loss variable PWR type reactor

International Nuclear Information System (INIS)

Yoshikuni, Masaaki.

1993-01-01

In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

International Nuclear Information System (INIS)

Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

2015-01-01

Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

An improved one-and-a-half group BWR core simulator for a new-generation core management system

International Nuclear Information System (INIS)

Iwamoto, Tatsuya; Yamamoto, Munenari

2000-01-01

An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

Human surfactant protein A2 gene mutations impair dimmer/trimer assembly leading to deficiency in protein sialylation and secretion.

Directory of Open Access Journals (Sweden)

Yi Song

Full Text Available Surfactant protein A2 (SP-A2 plays an essential role in surfactant metabolism and lung host defense. SP-A2 mutations in the carbohydrate recognition domain have been related to familial pulmonary fibrosis and can lead to a recombinant protein secretion deficiency in vitro. In this study, we explored the molecular mechanism of protein secretion deficiency and the subsequent biological effects in CHO-K1 cells expressing both wild-type and several different mutant forms of SP-A2. We demonstrate that the SP-A2 G231V and F198S mutants impair the formation of dimmer/trimer SP-A2 which contributes to the protein secretion defect. A deficiency in sialylation, but not N-linked glycosylation, is critical to the observed dimmer/trimer impairment-induced secretion defect. Furthermore, both mutant forms accumulate in the ER and form NP-40-insoluble aggregates. In addition, the soluble mutant SP-A2 could be partially degraded through the proteasome pathway but not the lysosome or autophagy pathway. Intriguingly, 4-phenylbutyrate acid (4-PBA, a chemical chaperone, alleviates aggregate formation and partially rescued the protein secretion of SP-A2 mutants. In conclusion, SP-A2 G231V and F198S mutants impair the dimmer/trimer assembly, which contributes to the protein sialylation and secretion deficiency. The intracellular protein mutants could be partially degraded through the proteasome pathway and also formed aggregates. The treatment of the cells with 4-PBA resulted in reduced aggregation and rescued the secretion of mutant SP-A2.

Spectral efficiency in crosstalk-impaired multi-core fiber links

Science.gov (United States)

Luís, Ruben S.; Puttnam, Benjamin J.; Rademacher, Georg; Klaus, Werner; Agrell, Erik; Awaji, Yoshinari; Wada, Naoya

2018-02-01

We review the latest advances on ultra-high throughput transmission using crosstalk-limited single-mode multicore fibers and compare these with the theoretical spectral efficiency of such systems. We relate the crosstalkimposed spectral efficiency limits with fiber parameters, such as core diameter, core pitch, and trench design. Furthermore, we investigate the potential of techniques such as direction interleaving and high-order MIMO to improve the throughput or reach of these systems when using various modulation formats.

Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

Energy Technology Data Exchange (ETDEWEB)

Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N. [Department of Astrophysical Sciences, 4 Ivy Lane, Princeton, NJ 08544 (United States); Coupon, Jean, E-mail: elinorm@astro.princeton.edu [Department of Astronomy, University of Geneva, ch. dEcogia 16, CH-1290 Versoix (Switzerland)

2017-02-10

There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

International Nuclear Information System (INIS)

Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N.; Coupon, Jean

2017-01-01

There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

Design of the core support and restraint structures for FFTF and CRBRP

International Nuclear Information System (INIS)

Sutton, H.G.; Rylatt, J.A.

1977-12-01

This paper presents and compares the design and fabrication of the FFTF and CRBRP reactor structures which support and restrain the reactor core assemblies. The fabrication of the core support structure (CSS) for the FFTF reactor was completed October 1972 and this paper discusses how the fabrication problems encountered with the FFTF were avoided in the subsequent design of the CRBR CSS. The radial core restraint structure of the FFTF was designed and fabricated such that an active system could replace the present passive system which is segmented and relies on the CSS core barrel for total structure integrity to maintain core geometry. The CRBR core restraint structure is designed for passive restraint only, and this paper discusses how the combined strengths of the restraint structure former rings and the CSS core barrel are utilized to maintain core geometry. Whereas the CSS for the FFTF interfaces directly with the reactor core assemblies, the CRBR CSS does not. A comparison is made on how intermediate structures in CRBR (inlet modules) provide the necessary design interfaces for supporting and providing flow distribution to the reactor core assemblies. A discussion is given on how the CRBR CSS satisfied the design requirements of the Equipment Specification, including thermal transient, dynamic and seismic loadings, and results of flow distribution testing that supported the CRBR design effort. The approach taken to simplify fabrication of the CRBR components, and a novel 20 inch deep narrow gap weld joint in the CSS are described

Modern X86 assembly language programming 32-bit, 64-bit, SSE, and AVX

CERN Document Server

Kusswurm, Daniel

2014-01-01

Modern X86 Assembly Language Programming shows the fundamentals of x86 assembly language programming. It focuses on the aspects of the x86 instruction set that are most relevant to application software development. The book's structure and sample code are designed to help the reader quickly understand x86 assembly language programming and the computational capabilities of the x86 platform. Major topics of the book include the following: 32-bit core architecture, data types, internal registers, memory addressing modes, and the basic instruction setX87 core architecture, register stack, special

Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

International Nuclear Information System (INIS)

Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

2011-01-01

The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

Modification of the ANC Nodal Code for analysis of PWR assembly bow

International Nuclear Information System (INIS)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

2008-01-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

Modification of the ANC Nodal Code for analysis of PWR assembly bow

Energy Technology Data Exchange (ETDEWEB)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

2008-07-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

2008-07-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

2009-04-15

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

International Nuclear Information System (INIS)

Shin, Hyeong-Ki

1999-01-01

The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

Neutronics aspects associated to irregular lattices in sodium fast reactors cores

International Nuclear Information System (INIS)

Gentili, Michele

2015-01-01

The fuel assemblies of SFR cores (sodium fast reactors) are normally arranged in hexagonal regular lattices, whose compactness is ensured in nominal operating conditions by thermal expansion of assemblies pads disposed on the six assembly wrapper faces. During the reactor operations, thermal expansion phenomena and irradiation creep phenomena occur and they cause the fuel assemblies to bow and to deform both radially and axially. The main goal of this PhD is the understanding of the neutronic aspects and phenomena occurring in case of core and lattice deformations, as much as the design and implementation of deterministic neutronic calculation schemes and methods in order to evaluate the consequences for the core design activities and the safety analysis. The first part of this work is focused on the development of an analytical model with the purpose to identify the neutronic phenomena that are the main contributors to the reactivity changes induced by lattice and core deformations. A first scheme based on the spatial mesh projection method has been conceived and implemented for the ERANOS codes (BISTRO, H3D and VARIANT) and to the SNATCH solver. The second calculation scheme propose is based on mesh deformation: the computing mesh is deformed as a function of the assembly displacement field. This methodology has been implemented for the solver SNATCH, which normally allows the Boltzmann equation to be solved for a regular mesh. Finally, an iterative method has been developed in order to fulfill an a-priori estimation of the maximal reactivity insertion as a function of the postulated mechanical energy provided to the core, as much as the deformation causing it. (author) [fr

Probing charge transfer dynamics in self-assembled monolayers by core hole clock approach

International Nuclear Information System (INIS)

Zharnikov, Michael

2015-01-01

This article reviews recent progress in the application of core hole clock approach in the framework of resonant Auger electron spectroscopy to the monomolecular assembles of alkyl, oligophenyl, and oligo(phenylene–ethynylene) based molecules on Au(1 1 1) substrates, referring mostly to the work by the author et al. The major goal was to study electron transfer (ET) dynamics in these systems serving as prototypes of molecular electronics (ME) devices. The ET pathway to the conductive substrate was unambiguously defined by resonant excitation of the nitrile tailgroup attached to the molecular backbone. Characteristic ET times within the femtosecond domain were determined, along with the attenuation factors for the ET dynamics, analogous to the case of the static transport. The above parameters were found to exhibit strong dependence on the character of the molecular orbital which mediates the ET process. In addition, certain spectral features, which can be associated with an inverse ET from the molecular backbone to the excitation site, were observed upon exchange of the nitrile group by strongly electronegative nitro moiety. The reported results represent a valuable input for theory and a certain potential for applications such as ME devices where optimization of ET can have significant technological impact.

Key aromatic/hydrophobic amino acids controlling a cross-amyloid peptide interaction versus amyloid self-assembly.

Science.gov (United States)

Bakou, Maria; Hille, Kathleen; Kracklauer, Michael; Spanopoulou, Anna; Frost, Christina V; Malideli, Eleni; Yan, Li-Mei; Caporale, Andrea; Zacharias, Martin; Kapurniotu, Aphrodite

2017-09-01

The interaction of the intrinsically disordered polypeptide islet amyloid polypeptide (IAPP), which is associated with type 2 diabetes (T2D), with the Alzheimer's disease amyloid-β (Aβ) peptide modulates their self-assembly into amyloid fibrils and may link the pathogeneses of these two cell-degenerative diseases. However, the molecular determinants of this interaction remain elusive. Using a systematic alanine scan approach, fluorescence spectroscopy, and other biophysical methods, including heterocomplex pulldown assays, far-UV CD spectroscopy, the thioflavin T binding assay, transmission EM, and molecular dynamics simulations, here we identified single aromatic/hydrophobic residues within the amyloid core IAPP region as hot spots or key residues of its cross-interaction with Aβ40(42) peptide. Importantly, we also find that none of these residues in isolation plays a key role in IAPP self-assembly, whereas simultaneous substitution of four aromatic/hydrophobic residues with Ala dramatically impairs both IAPP self-assembly and hetero-assembly with Aβ40(42). Furthermore, our experiments yielded several novel IAPP analogs, whose sequences are highly similar to that of IAPP but have distinct amyloid self- or cross-interaction potentials. The identified similarities and major differences controlling IAPP cross-peptide interaction with Aβ40(42) versus its amyloid self-assembly offer a molecular basis for understanding the underlying mechanisms. We propose that these insights will aid in designing intervention strategies and novel IAPP analogs for the management of type 2 diabetes, Alzheimer's disease, or other diseases related to IAPP dysfunction or cross-amyloid interactions. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

Evaluation of a thermal SCWR core with sub-channel analysis

International Nuclear Information System (INIS)

Liu Xiaojing; Cheng Xu

2008-01-01

A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Koyama, Jun-ichi; Ishibashi, Yoko; Mochida, Takaaki; Soneda, Hideo.

1994-01-01

In a fuel assembly having moderator rods, an axial average value of a ratio between the total of the lateral cross sectional area of a portion to be filled with moderators and the total of the lateral cross sectional area of fuel pellets is determined as greater than 0.4, a lateral cross sectional area of a portion to be filled with moderators per one moderator rod is determined as from 14 to 50cm 2 and the ratio between the total of the lateral cross sectional area of moderators and a total of the lateral cross sectional area of fuel pellets in a horizontal cross section is determined as from 2.7 to 3.4. Since the axial average value for lateral cross sectional area of a portion to be filled with moderators/lateral cross sectional area of fuel pellets is determined as ≥ 0.4, the lateral cross sectional area of moderators of moderator rods is increased, the lateral cross sectional area of a gap water region is decreased to reduce the value of local power peaking coefficient, so that thermal margin is ensured. At least one of the moderating rods is formed as a double-walled water rod tube to enhance an effect of spectral shift by flow rate control, reduce an uranium enrichment degree, and conduct operation without inserting control rods. (N.H.)

Core design characteristics of the hyper system

International Nuclear Information System (INIS)

Yonghee, Kim; Won-Seok, Park; Hill, R.N.

2003-01-01

In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1000 MWth lead-bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a ductless fuel assembly containing transuranics (TRU) dispersion fuel pins. In HYPER, a relatively high core height, 160 cm, is adopted to maximize the multiplication efficiency of the external source. In the ductless fuel assembly, 13 non-fuel rods are used as tie rods to maintain the mechanical integrity of assembly. As the reflector material, pure lead is used to improve the neutron economy and to minimise the generation of radioactive materials. In HYPER, to minimise the burn-up reactivity swing, a B 4 C burnable absorber is employed. For efficient depletion of the B-10 absorber, the burnable absorber is loaded only in the axially-central part (92 cm long) of the 13 tie rods of each assembly. In the current design, the amount of the B 4 C absorber was determined such that the burn-up reactivity swing is about 3.0% Δk. The long-lived fission products (LLFPs) 99 Tc and 129 I are also transmuted in the HYPER core such that their supporting ratios are equal to that of the TRUs. A heterogeneous LLFP transmutation in the reflector zone has been analysed in this work. A unique feature of the HYPER system is that it has an auxiliary core shutdown system, independent of the accelerator shutdown system. It has been shown that a cylindrical B 4 C absorber between the target and fuel blanket can drastically reduce the fission power even without shutting off the accelerator power. (author)

Effect of blanket assembly shuffling on LMR neutronic performance

International Nuclear Information System (INIS)

Khalil, H.; Fujita, E.K.

1987-01-01

Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

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).

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)

Whole Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor Considering the Gamma Energy Transport

International Nuclear Information System (INIS)

Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji

2012-01-01

Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design

Preliminary considerations on the startup phase for the ASTRID core

International Nuclear Information System (INIS)

Mignot, G.

2015-01-01

This paper presents preliminary considerations on the startup phase for the ASTRID core, as well as an overview of the different steps before reaching the optimised equilibrium core. The start-up phase is assumed to cover the period between loading the dummy core into the reactor (for commissioning tests) and achieving the optimised equilibrium core. Four main stages are considered: a first stage of start-up tests before fuel core loading, a second stage related to zero power and power ramp-up tests, a third stage corresponding to the transition from the first core to the equilibrium contractual core, and the last stage to reach the optimised performance for the equilibrium core. In the two last stages, a sub-assembly surveillance plan based on post-irradiation examinations is taken into account. As this work is in its preliminary stages, the first scenarios shown for the start-up phase must not be considered as the ASTRID reference scenarios. The scenarios strongly depend on the assumptions considered in the analysis, whereas those discussed in this paper aim at outlining the content and the duration of the starting phases for the ASTRID core, which will be useful in subsequently assessing the core sub-assembly fabrication needs. Assumptions for the start-up phase will be updated in accordance with progress on the ASTRID core design development and core qualification programme. (author)

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

International Nuclear Information System (INIS)

Londen, S.O.

1966-01-01

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

Energy Technology Data Exchange (ETDEWEB)

Londen, S O

1966-01-15

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important.

Determination of mixing factors for VVER-440 fuel assembly head

Energy Technology Data Exchange (ETDEWEB)

Tóth, S., E-mail: toth@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary); Aszódi, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary)

2013-11-15

CFD models have been developed for the heads of the old, the present and the new type VVER-440 fuel assemblies using the experience of a former validation process. With these models temperature distributions are investigated in the heads of some typical assemblies and the in-core thermocouple signals are calculated. The analyses show that the coolant mixing is intensive but not-perfect in the assembly heads. The difference between the thermocouple signal and the cross-sectional average temperature at the measurement level depends on the assembly type. Using the results of these CFD calculations the weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors the thermocouple signals are estimated and the results are statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that the deviations between the measured and the calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors.

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

2002-01-01

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

An integrated software system for core design and safety analyses: Cascade-3D

International Nuclear Information System (INIS)

Wan De Velde, A.; Finnemann, H.; Hahn, T.; Merk, S.

1999-01-01

The new Siemens program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of the most advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management. (authors)

In-core nuclear fuel management optimization of VVER1000 using perturbation theory

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser

2011-01-01

In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

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.

Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

International Nuclear Information System (INIS)

Hanan, N. A.; Deen, J.R.; Matos, J.E.

2005-01-01

Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

The integrated code system CASCADE-3D for advanced core design and safety analysis

International Nuclear Information System (INIS)

Neufert, A.; Van de Velde, A.

1999-01-01

The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

SMART core protection system design

International Nuclear Information System (INIS)

Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

2003-01-01

SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

Development of Core Design Technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

2007-06-15

This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Core@shell Nanoparticles: Greener Synthesis Using Natural Plant Products

Directory of Open Access Journals (Sweden)

Mehrdad Khatami

2018-03-01

Full Text Available Among an array of hybrid nanoparticles, core-shell nanoparticles comprise of two or more materials, such as metals and biomolecules, wherein one of them forms the core at the center, while the other material/materials that were located around the central core develops a shell. Core-shell nanostructures are useful entities with high thermal and chemical stability, lower toxicity, greater solubility, and higher permeability to specific target cells. Plant or natural products-mediated synthesis of nanostructures refers to the use of plants or its extracts for the synthesis of nanostructures, an emerging field of sustainable nanotechnology. Various physiochemical and greener methods have been advanced for the synthesis of nanostructures, in contrast to conventional approaches that require the use of synthetic compounds for the assembly of nanostructures. Although several biological resources have been exploited for the synthesis of core-shell nanoparticles, but plant-based materials appear to be the ideal candidates for large-scale green synthesis of core-shell nanoparticles. This review summarizes the known strategies for the greener production of core-shell nanoparticles using plants extract or their derivatives and highlights their salient attributes, such as low costs, the lack of dependence on the use of any toxic materials, and the environmental friendliness for the sustainable assembly of stabile nanostructures.

Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

Tablet—next generation sequence assembly visualization

Science.gov (United States)

Milne, Iain; Bayer, Micha; Cardle, Linda; Shaw, Paul; Stephen, Gordon; Wright, Frank; Marshall, David

2010-01-01

Summary: Tablet is a lightweight, high-performance graphical viewer for next-generation sequence assemblies and alignments. Supporting a range of input assembly formats, Tablet provides high-quality visualizations showing data in packed or stacked views, allowing instant access and navigation to any region of interest, and whole contig overviews and data summaries. Tablet is both multi-core aware and memory efficient, allowing it to handle assemblies containing millions of reads, even on a 32-bit desktop machine. Availability: Tablet is freely available for Microsoft Windows, Apple Mac OS X, Linux and Solaris. Fully bundled installers can be downloaded from http://bioinf.scri.ac.uk/tablet in 32- and 64-bit versions. Contact: tablet@scri.ac.uk PMID:19965881

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

Effects of low heterogeneity in fast critical assemblies

International Nuclear Information System (INIS)

Belov, S.P.; Dulin, V.A.; Zhukov, A.V.; Kuzin, E.N.; Mozhaev, V.K.; Sitnikov, V.I.; Tsibulya, A.M.; Shapar', A.V.; Zayfert, E.; Kuntsman, B.; Khayntsel'man, B.

1989-01-01

The problem of the low heterogeneity of fast critical assemblies, which are used to simulate fast reactors that are under design, has begun to assume increasing importance as the errors in nuclear data and group constants decrease and the capabilities of design codes improve. The design of the fuel channels of the fast critical assemblies of a BFS differs from that of the fuel subassemblies of a power reactor. The principal difference is that critical assemblies have a more heterogeneous structure than a reactor core does. As a result, the effects of this heterogeneity turn out to be appreciable for a number of functionals. Of particular interest was the measurement of the main neutronic characteristics of a fast reactor in its actual design and in the mockup produced by using BFS facilities. The authors measured and calculated the most important functionals (the ratios of the average cross sections of the main absorbing and fissioning elements, etc.) for both a homogeneous medium (fuel assemblies) and a heterogeneous medium (slugs, tubes) of practically identical composition. The objective of this work was to compare the discrepancy between experiment and calculations for the central functionals in the homogeneous and heterogeneous cases after corrections. This is a check of the accuracy of the simulation of homogeneous cores in fast power reactors by using the tools of the BFS fast critical assembly

Spring unit especially intended for a nuclear reactor core

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, Wilhelm.

1977-01-01

This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

Core arrangement in BWR type reactors

International Nuclear Information System (INIS)

Asano, Masayuki.

1981-01-01

Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

Interchange core/shell assembly of diluted magnetic semiconductor CeO2 and ferromagnetic ferrite Fe3O4 for microwave absorption

Directory of Open Access Journals (Sweden)

Jiaheng Wang

2017-05-01

Full Text Available Core/shell-structured CeO2/Fe3O4 and Fe3O4/CeO2 nanocapsules are prepared by interchange assembly of diluted magnetic semiconductor CeO2 and ferromagnetic ferrite Fe3O4 as the core and the shell, and vice versa, using a facile two-step polar solvothermal method in order to utilize the room-temperature ferromagnetism and abundant O-vacancies in CeO2, the large natural resonance in Fe3O4, and the O-vacancy-enhanced interfacial polarization between CeO2 and Fe3O4 for new generation microwave absorbers. Comparing to Fe3O4/CeO2 nanocapsules, the CeO2/Fe3O4 nanocapsules show an improved real permittivity of 3–10% and an enhanced dielectric resonance of 1.5 times at 15.3 GHz due to the increased O-vacancy concentration in the CeO2 cores of larger grains as well as the O-vacancy-induced enhancement in interfacial polarization between the CeO2 cores and the Fe3O4 shells, respectively. Both nanocapsules exhibit relatively high permeability in the low-frequency S and C microwave bands as a result of the bi-magnetic core/shell combination of CeO2 and Fe3O4. The CeO2/Fe3O4 nanocapsules effectively enhance permittivity and permeability in the high-frequency Ku band with interfacial polarization and natural resonance at ∼15 GHz, thereby improving absorption with a large reflection loss of -28.9 dB at 15.3 GHz. Experimental and theoretical comparisons with CeO2 and Fe3O4 nanoparticles are also made.

Research Programme for the 660 Mev Proton Accelerator Driven MOX-Plutonium Subcritical Assembly

CERN Document Server

Barashenkov, V S; Buttseva, G L; Dudarev, S Yu; Polanski, A; Puzynin, I V; Sissakian, A N

2000-01-01

The paper presents a research programme of the Experimental Acclerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton acceletator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO_2 + 75% UO_2) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k_eff = 0.945, energetic gain G = 30 and the accelerator beam power 0.5 kW.

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

Electrostatic Self-Assembly of Au Nanoparticles onto Thermosensitive Magnetic Core-Shell Microgels for Thermally Tunable and Magnetically Recyclable Catalysis.

Science.gov (United States)

Liu, Guoqiang; Wang, Daoai; Zhou, Feng; Liu, Weimin

2015-06-01

A facile route to fabricate a nanocomposite of Fe3O4@poly[N-isopropylacrylamide (NIPAM)-co-2-(dimethylamino)ethyl methacrylate (DMAEMA)]@Au (Fe3O4@PND@Au) is developed for magnetically recyclable and thermally tunable catalysis. The negatively charged Au nanoparticles with an average diameter of 10 nm are homogeneously loaded onto positively charged thermoresponsive magnetic core-shell microgels of Fe3O4@poly(NIPAM-co-DMAEMA) (Fe3O4@PND) through electrostatic self-assembly. This type of attachment offers perspectives for using charged polymeric shell on a broad variety of nanoparticles to immobilize the opposite-charged nanoparticles. The thermosensitive PND shell with swollen or collapsed properties can be as a retractable Au carrier, thereby tuning the aggregation or dispersion of Au nanoparticles, which leads to an increase or decrease of catalytic activity. Therefore, the catalytic activity of Fe3O4@PND@Au can be modulated by the volume transition of thermosensitive microgel shells. Importantly, the mode of tuning the aggregation or dispersion of Au nanoparticles using a thermosensitive carrier offers a novel strategy to adjust and control the catalytic activity, which is completely different with the traditional regulation mode of controlling the diffusion of reactants toward the catalytic Au core using the thermosensitive poly(N-isopropylacrylamide) network as a nanogate. Concurrent with the thermally tunable catalysis, the magnetic susceptibility of magnetic cores enables the Fe3O4@PND@Au nanocomposites to be capable of serving as smart nanoreactors for thermally tunable and magnetically recyclable catalysis. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Physics measurements on WWER-440 diagnostic assemblies

International Nuclear Information System (INIS)

Dach, K.; Jirousek, V.; Kott, J.; Horak, J.; Teren, S.; Nemec, J.

1980-01-01

The aims of physics measurements using diagnostic assemblies are the development of neutron noise diagnostics methods, the improvement of knowledge of the physical properties of the WWER reactor cores, the testing of computer programs, and the specification of nuclear safety criteria and the obtaining of information allowing the optimum nuclear fuel economy. The instrumentation of diagnostic assemblies is briefly described, including miniature fission chambers, SPN detectors and calorimeters. The method of evaluating and experimental testing is shown. (M.S.)

Alternating current dielectrophoresis of core-shell nanoparticles: Experiments and comparison with theory

Science.gov (United States)

Yang, Chungja

Nanoparticles are fascinating where physical and optical properties are related to size. Highly controllable synthesis methods and nanoparticle assembly are essential for highly innovative technological applications. Well-defined shaped and sized nanoparticles enable comparisons between experiments, theory and subsequent new models to explain experimentally observed phenomena. Among nanoparticles, nonhomogeneous core-shell nanoparticles (CSnp) have new properties that arise when varying the relative dimensions of the core and the shell. This CSnp structure enables various optical resonances, and engineered energy barriers, in addition to the high charge to surface ratio. Assembly of homogeneous nanoparticles into functional structures has become ubiquitous in biosensors (i.e. optical labeling), nanocoatings, and electrical circuits. Limited nonhomogenous nanoparticle assembly has only been explored. Many conventional nanoparticle assembly methods exist, but this work explores dielectrophoresis (DEP) as a new method. DEP is particle polarization via non-uniform electric fields while suspended in conductive fluids. Most prior DEP efforts involve microscale particles. Prior work on core-shell nanoparticle assemblies and separately, nanoparticle characterizations with dielectrophoresis and electrorotation, did not systematically explore particle size, dielectric properties (permittivity and electrical conductivity), shell thickness, particle concentration, medium conductivity, and frequency. This work is the first, to the best of our knowledge, to systematically examine these dielectrophoretic properties for core-shell nanoparticles. Further, we conduct a parametric fitting to traditional core-shell models. These biocompatible core-shell nanoparticles were studied to fill a knowledge gap in the DEP field. Experimental results (chapter 5) first examine medium conductivity, size and shell material dependencies of dielectrophoretic behaviors of spherical CSnp into 2D and

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

Capacitive short circuit detection in transformer core laminations

International Nuclear Information System (INIS)

Schulz, Carl A.; Duchesne, Stephane; Roger, Daniel; Vincent, Jean-Noel

2008-01-01

A capacitive measurement procedure is proposed that serves to detect burr-induced short circuits in transformer core laminations. The tests are conducted on stacks of transformer steel sheets as used for transformer core production and yield a short-circuit probability indicative of the additional eddy current losses to be expected. Applied during the assembly of transformer cores, the measurements can help to decide whether the burr treatment process is working efficiently or has to be readjusted

MOX fuel assembly and reactor core

International Nuclear Information System (INIS)

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

1998-01-01

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

Localization of adenovirus morphogenesis players, together with visualization of assembly intermediates and failed products, favor a model where assembly and packaging occur concurrently at the periphery of the replication center.

Directory of Open Access Journals (Sweden)

Gabriela N Condezo

2017-04-01

Full Text Available Adenovirus (AdV morphogenesis is a complex process, many aspects of which remain unclear. In particular, it is not settled where in the nucleus assembly and packaging occur, and whether these processes occur in a sequential or a concerted manner. Here we use immunofluorescence and immunoelectron microscopy (immunoEM to trace packaging factors and structural proteins at late times post infection by either wildtype virus or a delayed packaging mutant. We show that representatives of all assembly factors are present in the previously recognized peripheral replicative zone, which therefore is the AdV assembly factory. Assembly intermediates and abortive products observed in this region favor a concurrent assembly and packaging model comprising two pathways, one for capsid proteins and another one for core components. Only when both pathways are coupled by correct interaction between packaging proteins and the genome is the viral particle produced. Decoupling generates accumulation of empty capsids and unpackaged cores.

Steady state thermal hydraulic analysis of LMR core using COBRA-K code

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

1997-02-01

A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

Optimization of burnable poison disposition for in-core fuel assemblies

International Nuclear Information System (INIS)

Zhong Wenfa; Luo Rong; Zhou Quan

1997-09-01

The optimization of the burnable poison disposition in the initial core loading of the 200 MW nuclear heating reactor (NHR-200), is studied. The mass fraction of the burnable poison is used as the control variable with the objective to minimize the power peaking factor. The flexible tolerance method is used to solve the nonlinear programming optimal problem. The optimization method can be used in reactor physics design, and get a new pattern of initial core which is of reference value. (2 refs., 8 figs., 1 tab.)

WIMS/PANTHER analysis of UO2/MOX cores using embedded super-cells

International Nuclear Information System (INIS)

Knight, M.; Bryce, P.; Hall, S.

2012-01-01

This paper describes a method of analysing PWR UO 2 MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

International Nuclear Information System (INIS)

Hanan, N.A.; Liaw, J.R.; Matos, J.E.

2005-01-01

A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

Electrochemical synthesis of CORE-shell magnetic nanowires

KAUST Repository

Ovejero, Jesú s G.; Bran, Cristina; Vidal, Enrique Vilanova; Kosel, Jü rgen; Morales, Marí a P.; Vazquez, Manuel

2015-01-01

(Fe, Ni, CoFe) @ Au core-shell magnetic nanowires have been synthesized by optimized two-step potentiostatic electrodeposition inside self-assembled nanopores of anodic aluminium templates. The optimal electrochemical parameters (e.g., potential

Comment on the in-core measurement in the WWER nuclear power plant

International Nuclear Information System (INIS)

Krett, V.; Dach, K.; Erben, O.

1985-01-01

The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

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

The differential effects of core stabilization exercise regime and conventional physiotherapy regime on postural control parameters during perturbation in patients with movement and control impairment chronic low back pain

Directory of Open Access Journals (Sweden)

Fernandes Svetlana

2010-05-01

Full Text Available Abstract Background The purpose of the present study was to examine the differential effect of core stability exercise training and conventional physiotherapy regime on altered postural control parameters in patients with chronic low back pain (CLBP. As heterogeneity in CLBP population moderates the effect of intervention on outcomes, in this study, interventions approaches were used based on sub-groups of CLBP. Methods This was an allocation concealed, blinded, sequential and pragmatic control trial. Three groups of participants were investigated during postural perturbations: 1 CLBP patients with movement impairment (n = 15, MI group randomized to conventional physiotherapy regime 2 fifteen CLBP patients with control impairment randomized to core stability group (CI group and 3 fifteen healthy controls (HC. Results The MI group did not show any significant changes in postural control parameters after the intervention period however they improved significantly in disability scores and fear avoidance belief questionnaire work score (P 0.8 after 8 weeks of core stability exercises for the adjusted p values. Postural control parameters of HC group were analyzed independently with pre and post postural control parameters of CI and MI group. This revealed the significant improvements in postural control parameters in CI group compared to MI group indicating the specific adaptation to the core stability exercises in CI group. Though the disability scores were reduced significantly in CI and MI groups (p Conclusions In this study core stability exercise group demonstrated significant improvements after intervention in ground reaction forces (Fz, Mz; g > 0.8 indicating changes in load transfer patterns during perturbation similar to HC group. Trial registration UTRN095032158-06012009423714

Rapsodie first core manufacture. 1. part: processing plant

International Nuclear Information System (INIS)

Masselot, Y.; Bataller, S.; Ganivet, M.; Guillet, H.; Robillard, A.; Stosskopf, F.

1968-01-01

This report is the first in a series of three describing the processes, results and peculiar technical problems related to the manufacture of the first core of the fast reactor Rapsodie. A detailed study of manufacturing processes(pellets, pins, fissile sub-assemblies), the associated testings (raw materials, processed pellets and pins, sub-assemblies before delivery), manufacturing facilities and improvements for a second campaign are described. (author) [fr

EXPERIMENTAL STUDY OF 3D SELF-ASSEMBLED PHOTONIC CRYSTALS AND COLLOIDAL CORE-SHELL SEMICONDUCTOR QUANTUM DOTS

Directory of Open Access Journals (Sweden)

Pham Thu Nga

2017-11-01

Full Text Available In this contribution we present an experimental study of 3D opal photonic crystals. The samples are opals constituted by colloidal silica spheres, realized with self-assembly technique. The sphere diameter is selected in order to obtain coupling of the photonic band gap with the emission from CdSe/ZnS colloidal quantum dots. The quantum dots infiltrated in the opals is expected to be enhanced or suppressed depending on the detection angle from the photonic crystal. The structural and optical characterization of the SiO2 opal photonic crystals are performed by field-emission scanning electron microscopy and reflectivity spectroscopy. Measurements performed on samples permits to put into evidence the influence of the different preparation methods on the optical properties. Study of self-activated luminescence of the pure opals is also presented. It is shown that the luminescence of the sample with QDs have original QD emission and not due to the photonic crystal structure. The optical properties of colloidal core-shell semiconductor quantum dots of CdSe/ZnS which are prepared in our lab will be mention.

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)

Nuclear reactor fuel element sub-assemblies

International Nuclear Information System (INIS)

Hill, G.D.; Trevalion, P.A.

1977-01-01

A fuel element sub-assembly for a liquid metal cooled fast reactor is described. It comprises a bundle of fuel pins enclosed by a tubular wrapper having a lower end journal for plugging into an upper aperture in a core supporting structure and a spike bar with an articulated bush for engaging a lower aperture in the core supporting structure. The articulated bush is retained on a spherical end portion of the spike bar by a pair of parallel retaining pins arranged transversely and disposed one each side of the spike bar. The pins are tubular and collapsible at a predetermined loading to enable the spherical end portion to pass between them. The articulated bush has an internal groove for engagement by a lifting grab, this groove being formed in a bore for receiving the spherical end portion of the spike bar. The construction lessens liability to rattling of the fuel element sub-assemblies and aids removal for replacement. (U.K.)

Core baffle for nuclear reactors

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1977-01-01

The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

Fuel Assembly Damping Summary

Energy Technology Data Exchange (ETDEWEB)

Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

measurement testing under flow are also briefly discussed. Fuel assembly damping is an essential parameter to determine fuel assembly dynamic behavior in operating or accidental core. Dry damping coefficient from the out-pile pluck testing was used for the accident analysis model in conservative and simplified manner. But, this is way lower than wet or under-flow damping.

Safeguards on MOX assemblies at LWRs

International Nuclear Information System (INIS)

Arenas Carrasco, J.; Koulikov, I.; Heinonen, O.J.; Arlt, R.; Grigoleit, K.; Clarke, R.; Swinhoe, M.

2000-01-01

Operating within the framework of the New Partnership Approach (NPA) for unirradiated MOX fuel assemblies in LWRs, the IAEA and EURATOM have gained experience in safeguarding 13 LWRs licensed to operate with MOX assemblies. In order to fulfil SIR requirements, verification methods and techniques capable of measuring MOX assemblies under water have been and are still being developed. These encompass both qualitative tests for the detection of plutonium (gross attribute tests) and quantitative tests for the measurement of the amount of plutonium (partial defect tests) and are based on gamma and neutron detection techniques. There are nine PWR and two BWR where the reactor and the spent fuel pond can be covered by the same surveillance device. These are Type I reactors where the reactor and the pond are located in the same hall. In these types of facilities relying on surveillance during the MOX refuelling is especially difficult at the BWRs due to the depth of the core pond. There are two PWR type facilities where the reactor and the spent fuel pond are located in different halls and cannot be covered by the same surveillance device (Type II). An open core camera has not been installed during refuelling and therefore indirect surveillance is currently used to survey MOX loading. Improvements are therefore required and are under consideration. After receipt at the facility, there are a few facilities which must keep the received fresh MOX fuel in wet storage, not only for a short period prior to refuelling, but for more than a year, until the next refuelling campaign. In these cases timely inspections for direct use fresh nuclear material require considerable inspection effort. Additionally, where human surveillance of core loading and finally core closure are necessary there is also a large demand for manpower. Either an agreement should be reached with the operators to delay the MOX loading until the end of the fuelling campaign, or alternative approaches should be

Evaluation of the harmonics and neutronic coupling in the sub-cores of the Brazilian `RESUCO` subcritical assembly; Avaliacoes dos harmonicos e do acoplamento neutronico entre dois subnucleos do conjunto subcritico `RESUCO`

Energy Technology Data Exchange (ETDEWEB)

Aquino Bezerra, A.F. de

1991-05-01

The present study evaluates the importance of the harmonic components in the thermal neutron flux distribution and evaluates as well the separation required for attaining neutronic decoupling in sub-cores in subcritical assemblies. The theoretical results are compared to experimental ones performed at the Brazilian natural uranium, light water RESUCO subcritical assembly. It is observed that the harmonics have a very important contributions to neutron flux. Furthermore, the neutronic decoupling is attained with the removal of five rows of fuel elements, corresponding to 27,5 cm of light water. (F.E.). 23 refs, 18 figs, 9 tabs.

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.

Calculation of local flow conditions in the lower core of a PWR with code-Saturne

International Nuclear Information System (INIS)

Fournier, Y.

2003-01-01

In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. A code specialized for calculations in tube bundles is used to calculate the flow inside the whole of the core, with a resolution at the assembly level. Still, it is necessary to obtain realistic entry conditions, and these depend on the flow in the downcomer and lower plenum. Also, the flow in the first stages of the core features 4 incoming jets per assembly, and requires a resolution much finer than that used for the whole core calculation. A series of calculations are thus run with our incompressible Navier-Stokes solver, Code-Saturne, using a classical Ranse turbulence model. The first calculations involve a detailed geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate, being pierced with close to 800 holes, cannot be realistically represented within a practical mesh size, so that a head loss model is used. The lower core itself requiring even more detail is also represented with head losses. We make full use of Code-Saturne's non conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Starting just under the lower core, the mesh is aligned with fuel rod assemblies, so that different types of assemblies can be represented through different head loss coefficients. These calculations yield steady-state or near steady-state results, which are compared to experimental data, and should be sufficient to yield realistic entry conditions for full core calculations at assembly width resolution, and beyond those mechanical strain calculations. We are also interested in more detailed flow conditions and fluctuations in the lower core area, so as to better quantify vibrational input. This requires a much higher resolution, which we limit

Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

International Nuclear Information System (INIS)

Broc, Daniel

2001-01-01

As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

Structure and experimental program for SNEAK 12 - an assembly for fast breeder safety experiments

International Nuclear Information System (INIS)

Helm, F.

1979-01-01

The critical assembly SNEAK 12 and its foreseen experimental program have the main purpose to check the validity of neutron physics calculational methods used in the analysis of accidental situations of fast breeder reactors. In the investigation of accidental courses configurations have to be considered, which are caused by assembly deformation and meltdown and which are characterized by irregular fuel and structural material arrangements with cavities and empty channels. The reactivity differences between the unperturbed core and a series of perturbed configurations have to be determined. The individual configurations have to be chosen in such a way, that the calculational methods for the different aspects of the accident sequence (formation of cavities and channels, relocation of fissile and fertile material and steel) can be tested one by one. Two different cores are foreseen: SNEAK 12A as a one-zone core with enriched uranium fuel and SNEAK 12B with a central test zone with plutonium-uranium mixed-oxide fuel surrounded by a driver zone of enriched uranium. The report describes these cores and their assemblies, and the experimental program is outlined

Design Basis of Core Components and their Realization in the frame of the EPR'sTM Core Component Development

International Nuclear Information System (INIS)

Schebitz, Florian; Mekmouche, Abdelhalim

2008-01-01

Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI TM RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR TM , the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR TM . (authors)

Research programme for the 660 MeV proton accelerator driven MOX-plutonium subcritical assembly

International Nuclear Information System (INIS)

Barashenkov, V.S.; Buttsev, V.S.; Buttseva, G.L.; Dudarev, S.Yu.; Polanski, A.; Puzynin, I.V.; Sissakyan, A.N.

2000-01-01

The paper presents a research programme of the Experimental Accelerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton accelerator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO 2 + 75% UO 2 ) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k eff = 0.945, energetic gain G=30 and the accelerator beam power 0.5 kW

The EFR project: core and fuel

International Nuclear Information System (INIS)

Francillon, E.; Barnes, D.W.; Pay, A.; Wehmann, U.

1991-01-01

The draft studies on EFR has beginning, in March 1988. The first part of the summary draft has consisted in the core and fuel domains to harmonize the different approaches used in national projects (SPX2-SNR2-CDFR). Rapidly, the core First Consistent Design has been defined with references to the anterior conceptions. Since this definition, studies have been engaged on the management (mean burnup amelioration) and on the conception (breeding gain, sodium void coefficient reduction). After a presentation of the basis options and on the general conception of the fuel assemblies we make a point on the core and the interfaces with the fuel cycle [fr

Progress of full MOX core design in ABWR

International Nuclear Information System (INIS)

Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

2000-01-01

Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

Pragmatic Deficits and Social Impairment in Children with ADHD

Science.gov (United States)

Staikova, Ekaterina; Gomes, Hilary; Tartter, Vivien; McCabe, Allyssa; Halperin, Jeffrey M.

2013-01-01

Background: Impaired social functioning has been well documented in individuals with attention-deficit/hyperactivity disorder (ADHD). Existing treatments for ADHD are effective for managing core symptoms, but have limited effectiveness at improving social skills, suggesting that social deficits in ADHD may not be directly related to core symptoms…

Magnetic latch trigger for inherent shutdown assembly

International Nuclear Information System (INIS)

Sowa, E.S.

1976-01-01

An inherent shutdown assembly for a nuclear reactor is provided. A neutron absorber is held ready to be inserted into the reactor core by a magnetic latch. The latch includes a magnet whose lines of force are linked by a yoke of material whose Curie point is at the critical temperature of the reactor at which the neutron absorber is to be inserted into the reactor core. The yoke is in contact with the core coolant or fissionable material so that when the coolant or the fissionable material increase in temperature above the Curie point the yoke loses its magnetic susceptibility and the magnetic link is broken, thereby causing the absorber to be released into the reactor core. 6 claims, 3 figures

Capturing haplotypes in germplasm core collections

Science.gov (United States)

Genomewide data sets of single nucleotide polymorphisms (SNPs) offer great potential to improve ex situ conservation. Two factors impede their use for producing core collections. First, due to the large number of SNPs, the assembly of collections that maximize diversity may be intractable using ex...

Nuclear fuel assembly for fast neutron reactors

International Nuclear Information System (INIS)

Ilyunin, V.G.; Murogov, V.M.; Troyanov, M.F.; Rinejskij, A.A.; Ustinov, G.G.; Shmelev, A.N.

1982-01-01

The fuel assembly of a fast reactor consists of fuel elements comprising sections with fissionable and breeding material and tubes with hollows designed for entrapping gaseous fission products. Tubes joining up to the said sections are divided in a middle and a peripheral group such that at least one of the tube groups is placed in the space behind the coolant inlet ports. The configuration above allows reducing internal overpressure in the fuel assembly, thus reducing the volume of necessary structural elements in the core. (J.B.)

WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells

Energy Technology Data Exchange (ETDEWEB)

Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

2012-07-01

This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

Elevations in core and muscle temperature impairs repeated sprint performance

DEFF Research Database (Denmark)

Drust, B.; Rasmussen, P.; Mohr, Magni

2005-01-01

on a cycle ergometer in normal (approximately 20 degrees C, control) and hot (40 degrees C, hyperthermia) environments. RESULTS: Completion of the intermittent protocol in the heat elevated core and muscle temperatures (39.5 +/- 0.2 degrees C; 40.2 +/- 0.4 degrees C), heart rate (178 +/- 11 beats min(-1...... metabolic fatigue agents and we, therefore, suggest that it may relate to the influence of high core temperature on the function of the central nervous system.......)), rating of perceived exertion (RPE) (18 +/- 1) and noradrenaline (38.9 +/- 13.2 micromol l(-1)) (all P

Implementation of a fast running full core pin power reconstruction method in DYN3D

International Nuclear Information System (INIS)

Gomez-Torres, Armando Miguel; Sanchez-Espinoza, Victor Hugo; Kliem, Sören; Gommlich, Andre

2014-01-01

Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions

Implementation of a fast running full core pin power reconstruction method in DYN3D

Energy Technology Data Exchange (ETDEWEB)

Gomez-Torres, Armando Miguel [Instituto Nacional de Investigaciones Nucleares, Department of Nuclear Systems, Carretera Mexico – Toluca s/n, La Marquesa, 52750 Ocoyoacac (Mexico); Sanchez-Espinoza, Victor Hugo, E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-vom-Helmhotz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Kliem, Sören; Gommlich, Andre [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

2014-07-01

Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

Investigation of two-phase flow instability under SMART-P core conditions

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Lee, Chung Chan

2005-01-01

An integral-type advanced light water reactor, named SMART-P, is being continuously studied at KAERI. The reactor core consists of hundreds of closed-channel type fuel assemblies with vertical upward flows. The upper and lower parts of the fuel assembly channels are connected to the common heads. The constant pressure drop imposed on the channel is responsible for the occurrence of density wave oscillations under local boiling and/or natural circulation conditions. The fuel assembly channel with oscillatory flow is highly susceptible to experience the CHF which may cause the fuel failure due to a sudden increase of the cladding temperature. Thus, prevention of the flow instability is an important criterion for the SMART-P core design. Experimental and analytical studies have been conducted in order to investigate the onset of flow instability (OFI) under SMART core conditions. The parallel channel oscillations were observed in a high pressure water-loop test facility. A linear stability analysis model in the frequency-domain was developed for the prediction of the marginal stability boundary (MSB) in the parallel boiling channels

Dissecting the sequential assembly and localization of intraflagellar transport particle complex B in Chlamydomonas.

Directory of Open Access Journals (Sweden)

Elizabeth A Richey

Full Text Available Intraflagellar transport (IFT, the key mechanism for ciliogenesis, involves large protein particles moving bi-directionally along the entire ciliary length. IFT particles contain two large protein complexes, A and B, which are constructed with proteins in a core and several peripheral proteins. Prior studies have shown that in Chlamydomonas reinhardtii, IFT46, IFT52, and IFT88 directly interact with each other and are in a subcomplex of the IFT B core. However, ift46, bld1, and ift88 mutants differ in phenotype as ift46 mutants are able to form short flagella, while the other two lack flagella completely. In this study, we investigated the functional differences of these individual IFT proteins contributing to complex B assembly, stability, and basal body localization. We found that complex B is completely disrupted in bld1 mutant, indicating an essential role of IFT52 for complex B core assembly. Ift46 mutant cells are capable of assembling a relatively intact complex B, but such complex is highly unstable and prone to degradation. In contrast, in ift88 mutant cells the complex B core still assembles and remains stable, but the peripheral proteins no longer attach to the B core. Moreover, in ift88 mutant cells, while complex A and the anterograde IFT motor FLA10 are localized normally to the transition fibers, complex B proteins instead are accumulated at the proximal ends of the basal bodies. In addition, in bld2 mutant, the IFT complex B proteins still localize to the proximal ends of defective centrioles which completely lack transition fibers. Taken together, these results revealed a step-wise assembly process for complex B, and showed that the complex first localizes to the proximal end of the centrioles and then translocates onto the transition fibers via an IFT88-dependent mechanism.

Simulation of the Long period Core Design for WH type of KHNP

Energy Technology Data Exchange (ETDEWEB)

Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2016-10-15

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

Simulation of the Long period Core Design for WH type of KHNP

International Nuclear Information System (INIS)

Jung, Ji-Eun; Moon, Sang-Rae

2016-01-01

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

A coupling model for the two-stage core calculation method with subchannel analysis for boiling water reactors

International Nuclear Information System (INIS)

Mitsuyasu, Takeshi; Aoyama, Motoo; Yamamoto, Akio

2017-01-01

Highlights: • A coupling model of the two-stage core calculation with subchannel analysis. • BWR fuel assembly parameters are assumed and verified. • The model was evaluated for heterogeneous problems. - Abstract: The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

Increasing the working efficiency of core lifter in a sectional turbobit

Energy Technology Data Exchange (ETDEWEB)

Shakhmaev, Z M; Kutepov, A I; Ganiev, G M

1969-01-01

Reasons for decreases of core recovery and decreases of drilling efficiency in core drilling are presented. One of the main faults with the turbobit is breakage of core lifter. As the length of the core lifter was increased, its strength was reduced. Use of ordinary tubing is not satisfactory with core lifters equipped with turbobits of 170 and 235 mm diam. The construction of lightweight alloy tubing (73 mm diam) for use with 235 mm turbobits is described. Tests showed that the lightweight core lifter assembly works satisfactorily. Schematic diagrams and descriptions of lightweight core lifters are supplied.

Fuel loading method to exchangeable reactor core of BWR type reactor and its core

International Nuclear Information System (INIS)

Koguchi, Kazushige.

1995-01-01

In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

Improved core protection calculator system algorithm

International Nuclear Information System (INIS)

Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

2009-01-01

Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

Fast reactor core monitoring device

International Nuclear Information System (INIS)

Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

1982-01-01

Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

Research on plasma core reactors

International Nuclear Information System (INIS)

Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

1977-01-01

Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

International Nuclear Information System (INIS)

Feghhi, Seyed Amir Hossein; Rezazadeh, Marzieh; Kadi, Yacine; ); Tenreiro, Claudio; Aref, Morteza; Gholamzadeh, Zohreh

2013-01-01

The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can he adopted because a high fissile production rate of 233 U converted from 232 Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2 % enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core. (author)

Computer supervision of the core outlet sodium temperatures of FBTR

International Nuclear Information System (INIS)

Boopathy, C.

1976-01-01

Safety monitoring of the fast breeder test reactor at Kalpakkam (India) is achieved by a CDPS-on-line dual computer system which is dedicated to plant supervision. The on-line subsystem scans and supervises all the 170 core thermocouple signals every second. Organisation of the reactor core instruments, supervision of mean sodium outlet temperature and mean temperature drop across the core, detection of plugging of a fuel assembly are explained. (A.K.)

Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

International Nuclear Information System (INIS)

Ko, Y.J.; Euh, D.J.; Youn, Y.J.; Chu, I.C.; Kwon, T.S.

2011-01-01

A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

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

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

International Nuclear Information System (INIS)

Shin, Andong; Choi, Yong Won

2016-01-01

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

International Nuclear Information System (INIS)

Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

2009-01-01

Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

International Nuclear Information System (INIS)

Lelong, Franck

2010-01-01

In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

Device for removing a spent reactor core instrument tube

International Nuclear Information System (INIS)

Watanabe, Shigeru; Tsuji, Teruaki.

1980-01-01

Purpose: To easily and exactly execute works for removing a used reactor core instrument tube to be mounted in a reactor core from the lattice space of the core or for charging the tube into the lattice of the core. Constitution: When fuel assembly is pulled out of a reactor core and a spent reactor core instrument tube is then bent and removed from the core at periodical inspection time, a lower gripping unit integral with an upper gripping unit and a bending unit is provided at the lower end of a hanging rope of a winch, and lowered to the reactor core. Then, the spent reactor core instrument tube is gripped by the upper and lower gripping units, the bending unit is operated, the spent reactor core instrument tube is bent, and the tube is then pulled upwardly by the winch to remove the tube. (Aizawa, K.)

Design Basis of Core Components and their Realization in the frame of the EPR's{sup TM} Core Component Development

Energy Technology Data Exchange (ETDEWEB)

Schebitz, Florian [AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Mekmouche, Abdelhalim [AREVA NP SAS, 10 rue Juliette Recamier, 69456 Lyon Cedex 06 (France)

2008-07-01

Rod Cluster Control Assemblies (RCCAs), Thimble Plug Assemblies (TPAs), Primary Neutron Sources (PNS) and Secondary Neutron Sources (SNS) are essential for the operation of a Nuclear Power Plant. Different functional requirements ask for different components and geometries. Therefore three different core components are used within the primary circuit: - The RCCA, which contains the absorber materials, is used to regulate and shut down the nuclear chain reaction. Under these demanding conditions different effects are determining the lifetime of the RCCA and in particular of the control rods. Several improvements like ion-nitriding of the cladding, lengthening of the bottom end plug, helium backfilling and reduction of the absorber diameter in the bottom part, which have already been introduced with the HARMONI{sup TM} RCCA, show a real improvement in terms of lifetime. - The TPAs are used at positions without RCCAs and neutron sources to limit the by-pass flow-rate in the fuel assembly guide tubes. The advanced TPA design results from a perfect combination of French and German design experience feedback. Benefits like homogenized hydraulic flow and improved manageability in terms of handling tools show the joined experience. - The neutron sources are used to enhance the flux level when the core is sub-critical so as to facilitate the core start-up control by the neutron flux detectors. Primary and secondary neutron sources are designed in a common way with reviewed and improved methodology. As there are different ways and conditions to operate core components, several designs are available. For the EPR{sup TM}, the best methods and products have been chosen. All chosen components contribute to an optimized and safe operation of the EPR{sup TM}. (authors)

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

2008-01-01

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

Energy Technology Data Exchange (ETDEWEB)

Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

2008-09-15

The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

α-synuclein assemblies sequester neuronal α3-Na+/K+-ATPase and impair Na+ gradient

Science.gov (United States)

Shrivastava, Amulya Nidhi; Redeker, Virginie; Fritz, Nicolas; Pieri, Laura; Almeida, Leandro G; Spolidoro, Maria; Liebmann, Thomas; Bousset, Luc; Renner, Marianne; Léna, Clément; Aperia, Anita; Melki, Ronald; Triller, Antoine

2015-01-01

Extracellular α-synuclein (α-syn) assemblies can be up-taken by neurons; however, their interaction with the plasma membrane and proteins has not been studied specifically. Here we demonstrate that α-syn assemblies form clusters within the plasma membrane of neurons. Using a proteomic-based approach, we identify the α3-subunit of Na+/K+-ATPase (NKA) as a cell surface partner of α-syn assemblies. The interaction strength depended on the state of α-syn, fibrils being the strongest, oligomers weak, and monomers none. Mutations within the neuron-specific α3-subunit are linked to rapid-onset dystonia Parkinsonism (RDP) and alternating hemiplegia of childhood (AHC). We show that freely diffusing α3-NKA are trapped within α-syn clusters resulting in α3-NKA redistribution and formation of larger nanoclusters. This creates regions within the plasma membrane with reduced local densities of α3-NKA, thereby decreasing the efficiency of Na+ extrusion following stimulus. Thus, interactions of α3-NKA with extracellular α-syn assemblies reduce its pumping activity as its mutations in RDP/AHC. PMID:26323479

Fuel assemblies for nuclear reactors

International Nuclear Information System (INIS)

Jabsen, F.S.

1979-01-01

In a nuclear fuel assembly, hollow guide posts protrude into a fuel assembly and fitting grill from a biased spring pad with a plunger that moves with the spring pad plugging one end of each of the guide posts. A plate on the end fitting grill that has a hole for fluid discharge partially plugs the other end of the guide post. Pressurized water coolant that fills the guide post volume acts as a shock absorber and should the reactor core receive a major seismic or other shock, the fuel assembly is compelled to move towards a pad depending from a transversely disposed support grid. The pad bears against the spring pad and the plunger progressively blocks the orifices provided by slots in the guide posts thus gradually absorbing the applied shock. After the orifice has been completely blocked, controlled fluid discharge continues through a hole coil spring cooperating in the attenuation of the shock. (author)

Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

2002-03-01

The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

Nonlinear seismic analysis of a graphite reactor core

International Nuclear Information System (INIS)

Laframboise, W.L.; Desmond, T.P.

1988-01-01

Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

The upgrade of integrity analysis module and the mechanical behavior evaluation for the assembly duct

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Lee, Dong Uk; Kim, Young Il [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-05-01

The high neutron fluxes and operating temperatures associated with KALIMER are inducing the important radiation damage phenomena, which can cause significant dimensional changes in the core components of the reactor.The thermo-mechanical analysis of the assembly ducts for KALIMER are mainly performed to evaluate the following items.1) change of reactivity. 2) force at pads on core assemblies. 3) withdrawal force at refueling. 4) loading and refueling deviation of assembly ducts. 5) bowing modes for control assembly. In this report, the model for the evaluation of reactivity change as well as the refueling model and the withdrawl force model are upgraded. And the reactivity change is considered as the most important parameter among the above items. Therefore, the sensitivity analyses mainly associated with reactivity change are carried out. As the results, the pad gap between the assembly ducts preliminary driven for keeping the (-) reactivity change. 9 refs., 24 figs., 2 tabs. (Author)

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2013-01-01

The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

Support tube of in-core instruments

International Nuclear Information System (INIS)

Suzumura, Takeshi; Saito, Shozo; Yasuda, Tetsuo; Shirosaki, Kiyotaka.

1975-01-01

Object: To permit satisfactory output measurement by preventing the bending of a in-core instrument tube within a reactor due to vibrations by means of a spring and thereby preventing mechanical damage of an adjacent fuel channel box. Structure: At a corner of a channel box of a fuel assembly, a in-core instrument tube is arranged along a channel box and has its surface provided with a plurality of removable leaf springs arranged in the direction of axis of the in-core instrument tube and each having an arcular tip. Thus, when the in-core instrument tube is inserted into the reactor, the arcular tip portions of the leaf springs are brought into plane contact with the corner of the channel box so that the in-core instrument tube is elastically supported on the channel box. Thus, there is no possibility of causing damage to the adjacent fuel channel box. (Kamimura, M.)

ROSA full-core and DNBR capabilities

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

2012-01-01

This paper presents the latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost two decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

Immobilization of Three-Mile Island core debris

International Nuclear Information System (INIS)

Welch, J.M.; Miller, R.L.; Flinn, J.E.

1983-01-01

The immobilization of Three-Mile Island core debris in iron-enriched basalt (IEB), a fused-cast nuclear waste form, was considered. The amount of zirconium clad UO 2 fuel assemblies that can be dissolved in IEB using the Zr to UO 2 ratio present in the core was bracketed between 25 and 30% at 1500 0 C. The factors controlling the rate of dissolution of fuel pellets and Inconel, a structural component of the core, were investigated. Since the UO 2 dissolved in IEB could be a valuable resource in the future, the recovery of uranium from IEB using conventional ore-dressing and leaching techniques was assessed

Spectrophotometric Evaluation of Polyetheretherketone (PEEK as a Core Material and a Comparison with Gold Standard Core Materials

Directory of Open Access Journals (Sweden)

Bogna Stawarczyk

2016-06-01

Full Text Available This study investigated the colorimetric properties of different veneering materials on core materials. Standardized specimens (10 mm × 10 mm × 1.5 mm reflecting four core (polyetheretherketone (PEEK, zirconia (ZrO2, cobalt–chromium–molybdenum alloy (CoCrMo, and titanium oxide (TiO2; thickness: 1.5 mm and veneering materials (VITA Mark II, IPS e.max CAD, LAVA Ultimate and VITA Enamic, all in shade A3; thickness: 0.5, 1.0, 1.5 and 2 mm, respectively were fabricated. Specimens were superimposed to assemblies, and the color was determined with a spectrophotometer (CieLab-System or a chair-side color measurement device (VITA EasyShade, respectively. Data were analyzed using three-, two-, and one-way ANOVA, a Chi2-test, and a Wilson approach (p < 0.05. The measurements with EasyShade showed A2 for VITA Mark II, A3.5 for VITA Enamic, B2 for LAVA Ultimate, and B3 for IPS e.max CAD. LabE-values showed significant differences between the tested veneering materials (p < 0.001. CieLab-System and VITA EasyShade parameters of the different assemblies showed a significant impact of core (p < 0.001, veneering material (p < 0.001, and thickness of the veneering material (p < 0.001. PEEK as core material showed comparable outcomes as compared to ZrO2 and CoCrMo, with respect to CieLab-System parameters for each veneering material. The relative frequency of the measured VITA EasyShade parameters regarding PEEK cores also showed comparable results as compared to the gold standard CoCrMo, regardless of the veneering material used.

Core design of a high breeding fast reactor cooled by supercritical pressure light water

Energy Technology Data Exchange (ETDEWEB)

Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

2016-01-15

Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

Measurement and analysis of reactivity worth of 237Np sample in cores of TCA and FCA

International Nuclear Information System (INIS)

Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro; Suzaki, Takenori; Saito, Masaki

2009-01-01

The reactivity worth of 22.87 grams of 237 Np oxide sample was measured and analyzed in seven uranium cores in the Tank-Type Critical Assembly (TCA) and two uranium cores in the Fast Critical Assembly (FCA) at the Japan Atomic Energy Agency. The TCA cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The FCA cores, XXI and XXV, provided a hard neutron spectrum of the fast reactor and a soft one of the resonance energy region, respectively. Analyses were carried out using the JENDL-3.3 nuclear data library with a Monte Carlo method for the TCA cores and a deterministic method for the FCA cores. The ratios of calculated to experimental (C/E) reactivity worth were between 0.97 and 0.91, and showed no apparent dependence on the neutron spectrum. (author)

The influence of reactor core parameters on effective breeding coefficient Keff

Institute of Scientific and Technical Information of China (English)