MOX - equilibrium core design and trial irradiation in KAPS - 1

International Nuclear Information System (INIS)

Pradhan, A.S.; Ray, Sherly; Kumar, A.N.; Parikh, M.V.

2006-01-01

Option of usage of MOX fuel bundles in the equilibrium core of Indian 220 MWe PHWRs on a regular basis has been studied. The design of the MOX bundle considered is MOX -7 with inner 7 elements with uranium and plutonium oxide MOX fuel and outer 12 elements with natural uranium fuel. The composition of the plutonium isotopes corresponds to that at about 6500 MWD/TeU burnup. Burnup optimization has been done such that operation at design rated power is possible while achieving the maximum average discharge burnup. Operation with the optimized burnup pattern will result in substantial saving of natural uranium bundles. To obtain feedback on the performance of MOX bundles prior to its large scale use about 50 MOX-7 bundles have been loaded in KAPS - 1 equilibrium core. Locations have been selected such that reactor should be operating at rated power without violating any constraints on channel bundle powers and also meeting the safety requirements. Burnup of interest also should be achieved in minimum period of time. The fissile plutonium content in the 50 MOX fuel bundles loaded is about 75.6 wt % . About 38 bundles out of the 50 bundles loaded have been already discharged and remaining bundles are still in the core. The maximum discharge burnup of the MOX bundles is about 12000 MWD/TeU. The performance of the MOX bundles were excellent and as per prediction. No MOX bundle is reported to be failed. (author)

From Russian weapons grade plutonium to MOX fuel

International Nuclear Information System (INIS)

Braehler, G.; Kudriavtsev, E.G.; Seyve, C.

1997-01-01

The April 1996, G7 Moscow Summit on nuclear matters provided a political framework for one of the most current significant challenges: ensuring a consistent answer to the weapons grade fissile material disposition issue resulting from the disarmament effort engaged by both the USA and Russia. International technical assessments have showed that the transformation of Weapons grade Plutonium in MOX fuel is a very efficient, safe, non proliferant and economically effective solution. In this regard, COGEMA and SIEMENS, have set up a consistent technical program properly addressing incineration of weapons grade plutonium in MOX fuels. The leading point of this program would be the construction of a Weapons grade Plutonium dedicated MOX fabrication plant in Russia. Such a plant would be based on the COGEMA-SIEMENS industrial capabilities and experience. This facility would be operated by MINATOM which is the partner for COGEMA-SIEMENS. MINATOM is in charge of coordination of the activity of the Russian research and construction institutes. The project take in account international standards for non-proliferation, safety and waste management. France and Germany officials reasserted this position during their last bilateral summits held in Fribourg in February and in Dijon in June 1996. MINATOM and the whole Russian nuclear community have already expressed their interest to cooperate with COGEMA-SIEMENS in the MOX field. This follows governmental-level agreements signed in 1992 by French, German and Russian officials. For years, Russia has been dealing with research and development on MOX fabrication and utilization. So, the COGEMA-SIEMENS MOX proposal gives a realistic answer to the management of weapons grade plutonium with regard to the technical, industrial, cost and schedule factors. (author)

LTA Physics Design: Description of All MOX Pin LTA Design

International Nuclear Information System (INIS)

Pavlovichev, A.M.

2001-01-01

In this document issued according to Work Release 02.P.99-1b the results of neutronics studies of > MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M

Linear thermal expansion, thermal diffusivity and melting temperature of Am-MOX and Np-MOX

International Nuclear Information System (INIS)

Prieur, D.; Belin, R.C.; Manara, D.; Staicu, D.; Richaud, J.-C.; Vigier, J.-F.; Scheinost, A.C.; Somers, J.; Martin, P.

2015-01-01

Highlights: • The thermal properties of Np- and Am-MOX solid solutions were investigated. • Np- and Am-MOX solid solutions exhibit the same linear thermal expansion. • The thermal conductivity of Am-MOX is about 10% higher than that of Np-MOX. • The melting temperatures of Np-MOX and Am-MOX are 3020 ± 30 K and 3005 ± 30 K, respectively. - Abstract: The thermal properties of Np- and Am-MOX solid solution materials were investigated. Their linear thermal expansion, determined using high temperature X-ray diffraction from room temperature to 1973 K showed no significant difference between the Np and the Am doped MOX. The thermal conductivity of the Am-MOX is about 10% higher than that of Np-MOX. The melting temperatures of Np-MOX and Am-MOX, measured using a laser heating self crucible arrangement were 3020 ± 30 K and 3005 ± 30 K, respectively

Is PPARα intron 7 G/C polymorphism associated with muscle strength characteristics in nonathletic young men?

Science.gov (United States)

Broos, S; Windelinckx, A; De Mars, G; Huygens, W; Peeters, M W; Aerssens, J; Vlietinck, R; Beunen, G P; Thomis, M A

2013-08-01

Peroxisome proliferator-activated receptor alpha (PPARα), a ligand-dependent transcription factor, regulates fatty acid metabolism in heart and skeletal muscle. The intron 7 G/C polymorphism (rs4253778) has been associated with athletic performance. The rare C-allele was predominant in power athletes, whereas the G-allele was more frequent in endurance athletes. In the present study, we investigated the association between this polymorphism and strength characteristics in nonathletic, healthy young adults (n = 500; age 24.2 ± 4.4 years). Knee torque was measured during concentric knee flexion and extension movements at 60°/s, 120°/s, and 240°/s during 3, 25, and 5 repetitions, respectively. Also, resistance to muscle fatigue (i.e. work last 20% repetitions/work first 20% repetitions *100) was calculated. Differences in knee strength phenotypes between GG homozygous individuals and C-allele carriers were analyzed. The polymorphism did not influence the ability to produce isometric or dynamic knee flexor or extensor peak torque during static or dynamic conditions in this population (0.23 < P < 0.95). Similar results were found for the endurance ratio, a measure for resistance to muscle fatigue. In conclusion, the PPARα intron 7 G/C polymorphism does not seem to influence strength characteristics in a nonathletic population. © 2011 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Characterization of un-irradiated MIMAS MOX fuel by Raman spectroscopy and EPMA

Science.gov (United States)

Talip, Zeynep; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Valot, Christophe; Vauchy, Romain; Jégou, Christophe

2018-02-01

In this study, Raman spectroscopy technique was implemented to characterize un-irradiated MIMAS (MIcronized - MASter blend) MOX fuel samples with average 7 wt.% Pu content and different damage levels, 13 years after fabrication, one year after thermal recovery and soon after annealing, respectively. The impacts of local Pu content, deviation from stoichiometry and self-radiation damage on Raman spectrum of the studied MIMAS MOX samples were assessed. MIMAS MOX fuel has three different phases Pu-rich agglomerate, coating phase and uranium matrix. In order to distinguish these phases, Raman results were associated with Pu content measurements performed by Electron Microprobe Analysis. Raman results show that T2g frequency significantly shifts from 445 to 453 cm-1 for Pu contents increasing from 0.2 to 25 wt.%. These data are satisfactorily consistent with the calculations obtained with Gruneisen parameters. It was concluded that the position of the T2g band is mainly controlled by Pu content and self-radiation damage. Deviation from stoichiometry does not have a significant influence on T2g band position. Self-radiation damage leads to a shift of T2g band towards lower frequency (∼1-2 cm-1 for the UO2 matrix of damaged sample). However, this shift is difficult to quantify for the coating phase and Pu agglomerates given the dispersion of high Pu concentrations. In addition, 525 cm-1 band, which was attributed to sub-stoichiometric structural defects, is presented for the first time for the self-radiation damaged MOX sample. Thanks to the different oxidation resistance of each phase, it was shown that laser induced oxidation could be alternatively used to identify the phases. It is demonstrated that micro-Raman spectroscopy is an efficient technique for the characterization of heterogeneous MOX samples, due to its low spatial resolution.

Performance of the MTR core with MOX fuel using the MCNP4C2 code

International Nuclear Information System (INIS)

Shaaban, Ismail; Albarhoum, Mohamad

2016-01-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

Verification of the shift Monte Carlo code with the C5G7 reactor benchmark

International Nuclear Information System (INIS)

Sly, N. C.; Mervin, B. T.; Mosher, S. W.; Evans, T. M.; Wagner, J. C.; Maldonado, G. I.

2012-01-01

Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)

Results of postirradiation examination of the in-pile blockage experiments MOL-7C/4 and MOL-7C/5

International Nuclear Information System (INIS)

Weimar, P.; Schleisiek, K.

1991-01-01

The Mol-7C in-pile local blockage experiments are performed in the BR-2 reactor at Mol, Belgium as a joint project of Kernforchungszentrum Karlsruhe (KfK) and Studiecentrum voor Kernenergie/Centre d'Etude de l'Energie Nuclearire-Mol. The main objective is to investigate the consequences of local cooling disturbances in liquid-metal-cooled reactor (LMR) fuel subassemblies. In the tests Mol-7C/4 and MOL-7C/5, fuel pins from KNK II are used with a burnup of 5 and 1.7%, respectively. An active central porous blockage is used to simulate the cooling disturbance. During irradiation, the blockage causes significant local damage, including melting of cladding and fuel. Extensive postirradiation examinations (PIE) are performed to investigate the extent of damage. In this paper a description and interpretation of results of the destructive PIE performed at the Hot Cells Laboratory at KfK is given, along with some conclusions related to LMR safety

Design of a reactor core in the Oma Full MOX-ABWR

International Nuclear Information System (INIS)

Hama, Teruo

1999-01-01

The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

MOX fuel reprocessing and recycling

International Nuclear Information System (INIS)

Guillet, J.L.

1990-01-01

This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

Mox fuel experience: present status and future improvements

International Nuclear Information System (INIS)

Blanpain, P.; Chiarelli, G.

2001-01-01

Up to December 2000, more than 1700 MOX fuel assemblies have been delivered by Framatome ANP/Fragema to 20 French, 2 Belgian and 3 German PWRs. More than 1000 MOX fuel assemblies have been delivered by Framatome ANP GmbH (formerly Siemens) to 11 German PWRs and BWRs and to 3 Swiss PWRs. Operating MOX fuel up to discharge burnups of about 45,000 MWd/tM is done without any penalty on core operating conditions and fuel reliability. Performance data for fuel and materials have been obtained from an outstanding surveillance program. The examinations have concluded that there have been no significant differences in MOX fuel assembly characteristics relative to UO 2 fuel. The data from these examinations, combined with a comprehensive out-of-core and in-core analytical test program on the current fuel products, are being used to confirm and upgrade the design models necessary for the continuing improvement of the MOX product. As MOX fuel has reached a sufficient maturity level, the short term step is the achievement of the parity between UO 2 and MOX fuels in the EdF French reactors. This involves a single operating scheme for both fuels with an annual quarter core reload type and an assembly discharge burnup goal of 52,000 MWd/tM. That ''MOX parity'' product will use the AFA-3G assembly structure which will increase the fuel rod design margins with regards to the end-of-life internal pressure criteria. But the fuel development objective is not limited to the parity between the current MOX and UO 2 products: that parity must remain guaranteed and the MOX fuel managements must evolve in the same way as the UO 2 ones. The goal of the MOX product development program underway in France is the demonstration over the next ten years of a fuel capable of reaching assembly burnups of 70,000 MWd/tM. (author)

MOX and UOX PWR fuel performances EDF operating experience

International Nuclear Information System (INIS)

Provost, Jean-Luc; Debes, Michel

2005-01-01

Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

Developments in MOX fuel pellet fabrication technology: Indian experience

International Nuclear Information System (INIS)

Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

1998-01-01

India is interested in mixed oxide (MOX) fuel technology for better utilisation of its nuclear fuel resources. In view of this, a programme involving MOX fuel design, fabrication and irradiation in research and power reactors has been taken up. A number of experimental irradiations in research reactors have been carried out and a few MOX assemblies of ''All Pu'' type have been loaded in our commercial BWRs at Tarapur. An island type of MOX fuel design is under study for use in PHWRs which can increase the burn-up of the fuel by more than 30% compared to natural UO 2 fuel. The MOX fuel pellet fabrication technology for the above purpose and R and D efforts in progress for achieving better fuel performance are described in the paper. The standard MOX fuel fabrication route involves mechanical mixing and milling of UO 2 and PuO 2 powders. After detailed investigations with several types of mixing and milling equipments, dry attritor milling has been found to be the most suitable for this operation. Neutron Coincident Counting (NCC) technique was found to be the most convenient and appropriate technique for quick analysis of Pu content in milled MOX powder and to know Pu mixing is homogenous or not. Both mechanical and hydraulic presses have been used for powder compaction for green pellet production although the latter has been preferred for better reproducibility. Low residue admixed lubricants have been used to facilitate easy compaction. The normal sintering temperature used in Nitrogen-Hydrogen atmosphere is between 1600 deg. C to 1700 deg. C. Low temperature sintering (LTS) using oxidative atmospheres such as carbon dioxide, Nitrogen and coarse vacuum have also been investigated on UO 2 and MOX on experimental scale and irradiation behaviour of such MOX pellets is under study. Ceramic fibre lined batch furnaces have been found to be the most suitable for MOX pellet production as they offer very good flexibility in sintering cycle, and ease of maintainability

A novel fluffy nanostructured 3D network of Ni(C7H4O5) for supercapacitors

International Nuclear Information System (INIS)

Chen, Qiulin; Lei, Shuijin; Chen, Lianfu; Deng, Peiqin; Xiao, Yanhe; Cheng, Baochang

2017-01-01

Highlights: • The fluffy 3D network of Ni(C 7 H 4 O 5 ) complex is firstly prepared on Ni foam. • The fluffy 3D network shows high areal capacitance and excellent cycle stability. • The fluffy network has large superior pseudocapacitive performance than the powder. • An asymmetric supercapacitor with high capacitance and energy density is assembled. - Abstract: Supercapacitors have raised considerable research interest in recent years due to their extensive potential application in next-generation energy storage. It is always of great importance to develop new electrode materials for supercapacitors so far. In this research, nickel gallate complex (Ni(C 7 H 4 O 5 )) nanostructures are successfully grown on nickel foam by a facile hydrothermal route, which can be directly used as the electrodes for supercapacitors. X-ray diffraction patterns show that the sample is amorphous. The scanning electron microscopy images reveal that the products consist of novel fluffy 3D network with a mass of fibers. The electrochemical measurements demonstrate that the prepared Ni(C 7 H 4 O 5 ) electrode possesses the specific capacitance of 3.688 F cm −2 (1229.3 F g −1 ) at a current density of 9 mA cm −2 (3 A g −1 ). It presents an excellent cycling stability with a capacitance retention of 87.9% after 5000 cycles even at a very high current density of 40 mA cm −2 . An asymmetric supercapacitor device is assembled using the Ni(C 7 H 4 O 5 ) sample as positive electrode and activated carbon as negative one. A high gravimetric capacitance of 71.4 F g −1 at a current density of 0.5 A g −1 can be achieved. The fabricated device delivers the highest energy density of 23.8 W h kg −1 at a power density of 388.2 W kg −1 with a voltage window of 1.55 V. This strategy should be extended to other organometallic compounds for supercapacitors.

The association between the 4G/5G polymorphism in the promoter of the plasminogen activator inhibitor-1 gene and extension of postsurgical calf vein thrombosis.

Science.gov (United States)

Ferrara, Filippo; Meli, Francesco; Raimondi, Francesco; Montalto, Salvatore; Cospite, Valentina; Novo, Giuseppina; Novo, Salvatore

2013-04-01

The objective of this study was to evaluate whether the presence of a plasminogen activator inhibitor type 1 (PAI-1) promoter polymorphism 4G/5G could significantly influence the proximal extension of vein thrombosis in spite of anticoagulant treatment in patients with calf vein thrombosis (CVT) following orthopaedic, urological and abdominal surgery. We studied 168 patients with CVT, who had undergone orthopaedic, urological and abdominal surgery, subdivided as follows: first, 50 patients with thrombosis progression; second, 118 patients without thrombosis progression. The 4G/5G polymorphism of the plasminogen activator inhibitor 1 was evaluated in all patients and in 70 healthy matched controls. We also studied PAI-1 activity in plasma. The presence of 4G/5G genotype was significantly increased in the group of patients with the extension of thrombotic lesions and was associated with an increase in CVT extension risk (odds ratio adjusted for sex 2.692; 95% confidence interval 1.302-4.702). Moreover, we observed a significant increase of PAI-1 plasma activity in patients with extension of thrombotic lesion vs. patients without extension (P=0.0001). Patients with 4G/5G genotype in the promoter of the plasminogen activator inhibitor - 1 gene present a higher risk of extension of thrombotic lesions.

The MOX Demonstration Facility - the stepping stone to commercial MOX production

International Nuclear Information System (INIS)

Macdonald, A.G.

1994-01-01

The paper provides an insight into MOX fuel and the economic benefits of its use in pressurized water reactors (PWRs). BNFL and AEA are collaborating in the design, construction and operation of a thermal MOX Demonstration Facility (MDF) on the AEA Windscale site in Cumbria. The process flowsheet and equipment employed in MDF are discussed and the special precautions required to handle plutonium bearing materials are highlighted. The process flowsheet includes the short binderless route which has been specially developed for use in MDF and results in fuel pellets with an homogeneous structure. MDF is the forerunner to the design and construction of a larger scale Sellafield MOX Plant and hence is the stepping-stone to commercial MOX production. (author)

Full MOX core design in ABWR

International Nuclear Information System (INIS)

Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

2003-01-01

Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

The transportation of PuO2 and MOX fuel and management of irradiated MOX fuel

International Nuclear Information System (INIS)

Dyck, H.P.; Rawl, R.; Durpel, L. van den

2000-01-01

Information is given on the transportation of PuO 2 and mixed-oxide (MOX) fuel, the regulatory requirements for transportation, the packages used and the security provisions for transports. The experience with and management of irradiated MOX fuel and the reprocessing of MOX fuel are described. Information on the amount of MOX fuel irradiated is provided. (author)

Effect of mixing state on criticality safety evaluation in MOX powder and additive

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Miyoshi, Yoshinori

2005-01-01

Criticality safety analyses are discussed in which MOX powder and additive (e.g. zinc-stearate) are mixed in a powder treatment process of MOX fuel fabrication. The multiplication factor k eff is largely affected by how they are mixed, i.e., how the density and volume change with the mixing. In general, k eff increases when MOX powder is mixed with zinc-stearate. However, plutonium content and density of MOX powder make a difference in the k eff 's changes. Especially, MOX powder with a higher plutonium content and a higher density is not always unsafe in terms of criticality if it is mixed with zinc-stearate. (author)

Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, T.

2001-01-01

Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

Thermal conductivity of heterogeneous LWR MOX fuels

Science.gov (United States)

Staicu, D.; Barker, M.

2013-11-01

It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

Design Studies of ''Island'' Type MOX Lead Test Assembly

Energy Technology Data Exchange (ETDEWEB)

Pavlovitchev, A.M.

2000-03-31

In this document the results of neutronics studies of <> type MOX LTA design are presented. The characteristics both for infinite MOX grids and for VVER-1000 core with 3 MOX LTAs are calculated. the neutronics parameters of MOX fueled core have been performed using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

Mox fuels recycling

International Nuclear Information System (INIS)

Gay, A.

1998-01-01

This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

PHASE EVOLUTION AND MICROWAVE DIELECTRIC PROPERTIES OF (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) CERAMICS WITH ULTRA-LOW SINTERING TEMPERATURES

Science.gov (United States)

Zhou, Di; Guo, Jing; Yao, Xi; Pang, Li-Xia; Qi, Ze-Ming; Shao, Tao

2012-11-01

The (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) ceramics were prepared via the solid state reaction method. The sintering temperature decreased almost linearly from 755°C for (Li0.5Bi0.5)WO4 to 560°C for (Li0.5Bi0.5)MoO4. When the x≤0.3, a wolframite solid solution can be formed. For x = 0.4 and x = 0.6 compositions, both the wolframite and scheelite phases can be formed from the X-ray diffraction analysis, while two different kinds of grains can be revealed from the scanning electron microscopy and energy-dispersive X-ray spectrometer results. High performance of microwave dielectric properties were obtained in the (Li0.5Bi0.5)(W0.6Mo0.4)O4 ceramic sintered at 620°C with a relative permittivity of 31.5, a Qf value of 8500 GHz (at 8.2 GHz), and a temperature coefficient value of +20 ppm/°C. Complex dielectric spectra of pure (Li0.5Bi0.5)WO4 ceramic gained from the infrared spectra were extrapolated down to microwave range, and they were in good agreement with the measured values. The (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) ceramics might be promising for low temperature co-fired ceramic technology.

Full MOX core for PWRs

International Nuclear Information System (INIS)

Puill, A.; Aniel-Buchheit, S.

1997-01-01

Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

Characteristics of MOX dissolution with silver mediated electrolytic oxidation method

Energy Technology Data Exchange (ETDEWEB)

Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

MOX dissolution with silver mediated electrolytic oxidation method is to be applied to the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is though to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid. In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed from the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO{sub 2} purging. (author)

Evaluation of the hemoglobin A1c-analyzer TOSOH HLC-723 G7.

Science.gov (United States)

Gremmels, Heinz-Detlef; Richter, Anja; Watzke, Ivonne

2003-01-01

The TOSOH HLC-723 G7 is a compact analyzer designed for the measurement of HbA1c under routine laboratory conditions. The analyzer has an automatic blood tube supply and positive sample identification. Samples are transported automatically via racks in a continuous-load mode, cap piercing is optional. Tests devoted to the assessment of reproducibility and accuracy of analytical results indicated that over a test period of 17 days, the intra-assay variation (CV) was 1.79%, and the inter-assay variation 2.60%, respectively. A comparison with the predecessor model G5 showed a very good correlation (r = 0.997, y = 1.0041x - 0.00172; n = 149). The presence of high triglyceride, bilirubin or urea concentrations in patient samples did not influence the analytical precision. The labile HbA1c fraction (L-A1c) is clearly separated during chromatography and thus does not compromise HbA1c analysis. With a protocol of 1.2 minutes, the TOSOH G7 is a very fast analyzer, designed for laboratories with a high throughput of samples.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

Energy Technology Data Exchange (ETDEWEB)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

2013-07-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

International Nuclear Information System (INIS)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

2013-01-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS

International Nuclear Information System (INIS)

Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

1983-01-01

Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

BWRs with MOx fuel

International Nuclear Information System (INIS)

Demaziere, C.

1999-01-01

Calculations has been performed for loading BWRs with pure MOx or UOx/MOx fuel. It seems to be possible to load MOx bundles in BWRs, since most of the core characteristics are comparable with the ones of a full UOx core. Nevertheless two main problems arise: The shutdown margin at BOC is lower than 1%, this requires to have a new design for the control rods in order to increase their efficiency - but the problem can also be solved by modifying the Pu quality. The cores with MOx fuel are slightly less stable, unfortunately the simple model applied does not allow giving an absolute value for the decay ratio but only allows comparing the stability with the full UOx core

Compound Heterozygosity for Hb Alperton (HBB: c.407C>T) and IVS-I-5 (G>C) (HBB: c.92+5G>C) Mutations Presenting as a Moderate Anemia in an Indian Family.

Science.gov (United States)

Godbole, Koumudi G; Ramachandran, Angelina; Karkamkar, Ashwini S; Dalal, Ashwin B

2018-04-13

While knowledge of HBB gene mutations is necessary for offering prenatal diagnosis (PND) of β-thalassemia (β-thal), a genotype-phenotype correlation may not always be available for rare variants. We present for the first time, genotype-phenotype correlation for a compound heterozygous status with IVS-I-5 (G>C) (HBB: c.92+5G>C) and HBB: c.407C>T (Hb Alperton) mutations on the HBB gene in an Indian family. Hb Alperton is a very rare hemoglobin (Hb) variant with scant published information about its clinical presentation, especially when accompanied with another HBB gene mutation. Here we provide biochemical as well as clinical details of this variant.

The MOX

International Nuclear Information System (INIS)

Legay, Christophe

1997-06-01

In this report, the author first proposes a presentation of plutonium with a brief history of its discovery and the discovery of other transuranic elements, a presentation of its main characteristics, and a description of its production ways. He also proposes an overview of data regarding world plutonium production and plutonium stock situation. The second part addresses the MOX fuel in relationship with the choice of non proliferation. The author describes the MOX fuel cycle (production, use in reactor, and reprocessing) and outlines the environmental and economic benefits of this fuel, and its interest within the frame of struggle against nuclear proliferation. The third part addresses the present situation and perspectives. He comments the American posture (principles and recent statements), discusses alternatives regarding nuclear wastes, and outlines MOX opportunities by evoking the French case and international perspectives, and the benefits in terms of matching irreversibility and safety

Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

International Nuclear Information System (INIS)

Ito, Chihiro

1999-01-01

It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

Transport of MOX fuel

International Nuclear Information System (INIS)

Porter, I.R.; Carr, M.

1997-01-01

The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

Development of MOX manufacturing technology in BNFL

International Nuclear Information System (INIS)

Buchan, P.G.; Powell, D.J.; Edwards, J.

1998-01-01

BNFL is successfully operating a small scale MOX fuel fabrication facility at its Sellafield Site and is currently constructing an advanced, commercial scale MOX facility to complement its existing LWR UO 2 fabrication capability. BNFL's MOX fuel capability is fully supported by a comprehensive technology development programme aimed at providing a high quality product which is successfully competing in the market. Building on the experience gained over the last 30 years, is from the production of both thermal and fast reactor MOX fuels, BNFL's development team set a standard for its MOX product which is targeted at exceeding the performance of UO 2 fuel in reactor. In order to meet the stringent design requirements the product development team has introduced the Short Binderless Route (SBR) process that is now used routinely in BNFL's MOX Demonstration Facility (MDF) and which forms the basis for BNFL's large scale Sellafield MOX Plant. This plant not only uses the SBR process for MOX production but also incorporates the most advanced technology available anywhere in the world for nuclear fuel production. A detailed account of the technology developed by BNFL to support its MOX fuels business will be provided, together with an explanation of the processes and plants used for MOX fuel production by BNFL. The paper also looks at the future needs of the MOX business and how improvements in pellet design can assist the MOX fabrication production process to meet the user demand requirements of utilities around the world. (author)

MOX fuel design and development consideration

International Nuclear Information System (INIS)

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

1997-01-01

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

Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

International Nuclear Information System (INIS)

Chakir, E.; Erradi, L.; Bardouni, T El.; Khoukhi, T El.; Boukhal, H.; Meroun, O.; Bakkari, B El

2007-01-01

Full text: In a previous work, we have analysed the main french experiments available on the reactivity temperature coefficient (RTC) : CREAOLE and Mistral experiments. In these experiments, the RTC has been measured in both UO2 and UO2-PuO2 PWR type lattices. Our calculations, using APPOLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO2 lattices is less than 1 pcm/Deg C which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges : an average error of -2 ± 0.5 pcm/Deg C is observed in low temperatures and an error of +3±2 pcm/Deg C is obtained for temperature higher than 250Deg C. In the present work, we analysed additional experimental benchmarks on the RTC of UO2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo Method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. Thus we have used the code MCNP5, for its recognized power and its availability. This analysis shows for the UO2 lattices, an average experiment-calculation deviation of about 0,5 pcm/Deg C, which is largely below the target accuracy for this type of lattices, that we estimate at approximately 1 pcm/Deg C. For the KAMINI experiment, which relates to the measurement of the RTC in light water moderated lattice using U-233 as fuel our analysis shows that the Endf/B6 library gives the best result, with an experiment -calculation deviation of the order of -0,16 pcm/Deg C. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7pcm/Deg C ( for a range of temperature going from 20 to 248 Deg C) and -1.2 pcm/Deg C ( for a range of temperature going from 20 to

MOx Depletion Calculation Benchmark

International Nuclear Information System (INIS)

San Felice, Laurence; Eschbach, Romain; Dewi Syarifah, Ratna; Maryam, Seif-Eddine; Hesketh, Kevin

2016-01-01

Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of Reactor Systems (WPRS) has been established to study the reactor physics, fuel performance, radiation transport and shielding, and the uncertainties associated with modelling of these phenomena in present and future nuclear power systems. The WPRS has different expert groups to cover a wide range of scientific issues in these fields. The Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS) was created in 2011 to perform specific tasks associated with reactor physics aspects of present and future nuclear power systems. EGRPANS provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for different reactor systems and also provides specific technical information regarding: core reactivity characteristics, including fuel depletion effects; core power/flux distributions; Core dynamics and reactivity control. In 2013 EGRPANS published a report that investigated fuel depletion effects in a Pressurised Water Reactor (PWR). This was entitled 'International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues' NEA/NSC/DOC(2013) that documented a benchmark exercise for UO 2 fuel rods. This report documents a complementary benchmark exercise that focused on PuO 2 /UO 2 Mixed Oxide (MOX) fuel rods. The results are especially relevant to the back-end of the fuel cycle, including irradiated fuel transport, reprocessing, interim storage and waste repository. Saint-Laurent B1 (SLB1) was the first French reactor to use MOx assemblies. SLB1 is a 900 MWe PWR, with 30% MOx fuel loading. The standard MOx assemblies, used in Saint-Laurent B1 reactor, include three zones with different plutonium enrichments, high Pu content (5.64%) in the center zone, medium Pu content (4.42%) in the intermediate zone and low Pu content (2.91%) in the peripheral zone

MOX use in PWRs. EDF operation experience

International Nuclear Information System (INIS)

Provost, Jean-Luc; Debes, Michel

2011-01-01

From the origin, EDF back-end fuel cycle strategy has focused on 'closing the fuel cycle', in other words integrating fuel reprocessing, with vitrification of high level waste concentrated within small volumes, and the recycling of valuable materials. The implementation of this policy was marked in 1987 by the first loading of sixteen MOX. By December 2010, 20 reactors have been loaded with 1750 tHM of MOX. EDF current strategy is to match the reprocessing program with MOX manufacturing capacity to limit the quantity of separated plutonium. This is routinely called the 'flow ad-equation' strategy. Currently, the MOX Parity core management achieves balance of MOX and UOX performance with a significant increase of the MOX discharge burn-up. Globally, the behavior under irradiation of MOX fuel assemblies has been satisfactory. So far, from the beginning of MOX use in EDF PWRs, only 6 MOX FAs with rod leakage have been identified, which gives a very satisfactory level of reliability. The industrial maturity of MOX fuel, with increased performances, allows the improvement of nuclear KWh competitiveness and of the plant operation performance, while maintaining in operation the same safety level, without significant impact on environment and radiological protection. (author)

On the thermal evolution of Pu-rich agglomerates in MOX

International Nuclear Information System (INIS)

Verwerft, M.; Leenaers, A.; Lippens, M.; Mertens, L.

1999-01-01

From the experience accumulated so far on irradiated MOX fuel, its overall behaviour under irradiation is generally well predicted by existing fuel models. It appears however that additional data are still welcome to properly benchmark fission gas release models, mainly at elevated burnup. To this aim, an international research project, FIGARO, was initiated. Its goal was to provide thermal and fission gas release data og MOX at high burnup. Two MOX fuel rods irradiated to high burnup (50 GWd/tM peak pellet) but at lower power (less than 200 W/cm) were selected for segmentation and instrumentation with central thermocouple and pressure gauge. The instrumented segments were subjected to irradiations at variable linear power in the HALDEN MTR. Both temperature and internal pressure were online monitored during the ramp test. Afterwards, the rod segments were transported and extensively investigated. The paper focuses on the investigation of the evolution of the microstructure of Pu-rich agglomerates as a function of temperature

Technology developments for Japanese BWR MOX fuel utilization

International Nuclear Information System (INIS)

Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

1997-01-01

The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

MOX fuel transport: the French experience

International Nuclear Information System (INIS)

Sanchis, H.; Verdier, A.; Sanchis, H.

1999-01-01

In the back-end of the fuel cycle, several leading countries have chosen the Reprocessing, Conditioning, Recycling (RCR) option. Plutonium recycling in the form of MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants an several European countries. The COGEMA Group has developed the industrialized products to master the RCR operation including transport COGEMA subsidiary, TRANSNUCLEAIRE has been operating MOX fuel transports on an industrial scale for more than 10 years. In 1998, around 200 transports of Plutonium materials have been organised by TRANSNUCLEAIRE. These transports have been carried out by road between various facilities in Europe: reprocessing plants, manufacturing plants and power plants. The materials transported are either: PuO 2 and MOX powder; BWR and PWR MOX fuel rods; BWR and PWR MOX fuel assemblies. Because MOX fuel transport is subject to specific safety, security and fuel integrity requirements, the MOX fuel transport system implemented by TRANSNUCLEAIRE is fully dedicated. Packaging have been developed, licensed and manufactured for each kind of MOX material in compliance with relevant regulations. A fleet of vehicles qualified according to existing physical protection regulations is operated by TRANSNUCLEAIRE. TRANSNUCLEAIRE has gained a broad experience in MOX transport in 10 years. Technical and operational know-how has been developed and improved for each step: vehicles and packaging design and qualification; vehicle and packaging maintenance; transport operations. Further developments are underway to increase the payload of the packaging and to improve the transport conditions, safety and security remaining of course top priority. (authors)

Design of full MOX core in ABWR

International Nuclear Information System (INIS)

Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

1999-01-01

A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

The MOX fuel in France

International Nuclear Information System (INIS)

2011-01-01

This document briefly describes the MOX production cycle which is performed in the MELOX plant in Marcoule by AREVA. It briefly indicates the main risks occurring during the whole MOX production and use cycle. They are associated with MOX production (high neutron and gamma dose rates, contamination, criticality, heat release), transportation, its use in reactors, its storage in pools after irradiation. All these stages need radiation protection measures

Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

International Nuclear Information System (INIS)

Erradi, L.; Chetaine, A.; Chakir, E.; Kharchaf, A.; Elbardouni, T.; Elkhoukhi, T.

2005-01-01

In a previous work, we have analysed the main French experiments available on the reactivity temperature coefficient (RTC): CREOLE and MISTRAL experiments. In these experiments, the RTC has been measured in both UO 2 and UO 2 -PuO 2 PWR type lattices. Our calculations, using APOLLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO 2 lattices is less than 1 pcm/C degrees which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges: an average error of -2 ± 0.5 pcm/C degrees is observed in low temperatures and an error of +3 ± 2 pcm/C degrees is obtained for temperatures higher than 250 C degrees. In the present work, we analysed additional experimental benchmarks on the RTC of UO 2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. This analysis shows for the UO 2 lattices, a maximum experiment-calculation deviation of about 0,7 pcm/C degrees, which is below the target accuracy for this type of lattices. For the KAMINI experiment, which relates to the measurement of the RTC in a light water moderated lattice using U-233 as fuel our analysis shows that the ENDF/B6 library gives the best result, with an experiment-calculation deviation of the order of -0,16 pcm/C degrees. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7 pcm/C degrees (for a range of temperatures going from 20 to 248 C degrees) and -1,2 pcm/C degrees (for a range of temperatures going from 20 to 80 C degrees). This result, in particular the tendency which has the error to decrease when the

Study on high performance MOX fuel and core design in full MOX ABWR(1) by GNF-J

International Nuclear Information System (INIS)

Izutsu, Sadayuki; Goto, Daisuke; Saeki, Jun; Kokubun, Takehiro; Yokoya, Jun

2003-01-01

The concepts of high-performance MOX fuel using 10x10 lattices suitable for full-MOX ABWR are shown in this paper, in which average discharge exposure is extended up to 45 GWd/t with heavy-metal inventory increased over current MOX, reducing the number of refueling bundles, resulting in fuel cycle cost reduction and core performance satisfaction. Also, the increase of Pu inventory is taken into account from the viewpoint to extend the flexibility of MOX fuel utilization. (author)

Transport of MOX fuel from Europe to Japan; Transport de combustible mox d' Europe vers le Japon

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

International Nuclear Information System (INIS)

Nagano, M.; Sakurai, S.; Yamaguchi, H.

1997-01-01

MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

MOX in reactors: present and future

International Nuclear Information System (INIS)

Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis

2010-01-01

In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

Inhibition of translation by 7-methyl guanosine (m7G) nucleotide cap analogs with derivatized 5'-monophosphates

International Nuclear Information System (INIS)

Tahara, S.M.; Darzynkiewicz, E.; Ekiel, I.

1986-01-01

Recognition of the 5'-m 7 GpppN (cap) structure of eukaryote mRNA is an important step of translation initiation as shown by the potent inhibitory effect of m 7 G nucleotides on this process. A comparison of cap analogs as competitive inhibitors of initiation has allowed the authors to map probable protein-ligand contact points between the cap and cognate cap binding proteins (CBPs). Recently, several new derivatives of m 7 GMP (1) with modified phosphates were synthesized: m 7 G 5'-phosphite (2), m 7 G 5'-phosphoramidate (3), m 7 G 5'-methylphosphonate (4), and m 7 G 5'-phosphate-O-methyl ester (5). In addition, 7,8-dimethyl GMP (6) and 7-methyl 8-amino GMP (7) were synthesized. 6 and7 are primarily syn and anti respectively, relative to the glycosidic bond as shown by solution NMR studies. Inhibition by analogs on total translation in reticulocyte lysate and binding of mRNA to rabbit reticulocyte ribosomes was found to be: 1 = 3 > 5 > 4 > 2. The inhibitory activity of 3 was unexpected since it is isosteric with 4, however it suggested that electron configuration and/or the ability to form a hydrogen bond between protein and the phosphate moiety might be important for ligand binding. 7 was more inhibitory than 6. The latter two are isosteric therefore differences in electron delocalization and/or syn-anti conformation are likely to be the reason(s) for the observed difference

Confirmation test of powder mixing process in J-MOX

International Nuclear Information System (INIS)

Ota, Hiroshi; Osaka, Shuichi; Kurita, Ichiro

2009-01-01

Japan Nuclear Fuel Ltd. (hereafter, JNFL) MOX Fuel Fabrication Plant (hereafter, J-MOX) is what fabricates MOX fuel for domestic light water power plants. Development of design concept of J-MOX was started mid 90's and the frame of J-MOX process was clarified around 2000 including adoption of MIMAS process as apart of J-MOX powder process. JNFL requires to take an answer to any technical question that has not been clarified ever before by world's MOX and/or Uranium fabricators before it commissions equipment procurement. J-MOX is to be constructed adjacent to the Rokkasho Reprocessing Plant (RRP) and to utilize MH-MOX powder recovered at RRP. The combination of the MIMAS process and the MH-MOX powder is what has never tried in the world. Therefore JNFL started a series of confirmation tests of which the most important is the powder test to confirm the applicability of MH-MOX powder to the MIMAS process. The MH-MOX powder, consisting of 50% plutonium oxide and 50% uranium oxide, originates JAEA development utilizing microwave heating (MH) technology. The powder test started with laboratory scale small equipment utilizing both uranium and the MOX powder in 2000, left a solution to tough problem such as powder adhesion onto equipment, and then was followed by a large scale equipment test again with uranium and the MOX powder. For the MOX test, actual size equipment within glovebox was manufactured and installed in JAEA plutonium fuel center in 2005, and based on results taken so far an understanding that the MIMAS equipment, with the MH-MOX powder, can present almost same quality MOX pellet as what is introduced as fabricated in Europe was developed. The test was finished at the end of Japanese fiscal year (JFY) 2007, and it was confirmed that the MOX pellets fabricated in this test were almost satisfied with the targeted specifications set for domestic LWR MOX fuels. (author)

Development of MOX fuel database

International Nuclear Information System (INIS)

Ikusawa, Yoshihisa; Ozawa, Takayuki

2007-03-01

We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

Determination of chloride in MOX samples using chloride ion selective electrode

Energy Technology Data Exchange (ETDEWEB)

Govindan, R; Das, D K; Mallik, G K; Sumathi, A; Patil, Sangeeta; Raul, Seema; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

1997-09-01

The chloride present in the MOX fuel is separated from the matrix by pyrohydrolysis at a temperature of 950 {+-} 50 degC and is then analyzed by chloride ion selective electrode (Cl-ISE). The range covered is 0.4-4 ppm with a precision of better than {+-}5% R.S.D. (author). 4 refs., 1 tab.

Program on MOX fuel utilization in light water reactors

International Nuclear Information System (INIS)

Kenda, Hirofumi

2000-01-01

MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

MOX Cross-Section Libraries for ORIGEN-ARP

International Nuclear Information System (INIS)

Gauld, I.C.

2003-01-01

The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

Transport of MOX fuel from Europe to Japan; Transport de combustible mox d' Europe vers le Japon

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

Fission gas release behaviour in MOX fuels

International Nuclear Information System (INIS)

Viswanathan, U.K.; Anantharaman, S.; Sahoo, K.C.

2002-01-01

As a part of plutonium recycling programme MOX (U,Pu)O 2 fuels will be used in Indian boiling water reactors (BWR) and pressurised heavy water reactors (PHWR). Based on successful test irradiation of MOX fuel in CIRUS reactor, 10 MOX fuel assemblies have been loaded in the BWR of Tarapur Atomic Power Station (TAPS). Some of these MOX fuel assemblies have successfully completed the initial target average burnup of ∼16,000 MWD/T. Enhancing the burnup target of the MOX fuels and increasing loading of MOX fuels in TAPS core will depend on the feedback information generated from the measurement of released fission gases. Fission gas release behaviour has been studied in the experimental MOX fuel elements (UO 2 - 4% PuO 2 ) irradiated in pressurised water loop (PWL) of CIRUS. Eight (8) MOX fuel elements irradiated to an average burnup of ∼16,000 MWD/T have been examined. Some of these fuel elements contained controlled porosity pellets and chamfered pellets. This paper presents the design details of the experimental set up for studying fission gas release behaviour including measurement of gas pressure, void volume and gas composition. The experimental data generated is compared with the prediction of fuel performance modeling codes of PROFESS and GAPCON THERMAL-3. (author)

MOX fuel for Indian nuclear power programme

International Nuclear Information System (INIS)

Kamath, H.S.; Anantharaman, K.; Purushotham, D.S.C.

2000-01-01

A sound energy policy and a sound environmental policy calls for utilisation of plutonium (Pu) in nuclear power reactors. The paper discusses the use of Pu in the form of mixed oxide (MOX) fuel in two Indian boiling water reactors (BWRs) at Tarapur. An industrial scale MOX fuel fabrication plant is presently operational at Tarapur which is capable of manufacturing MOX fuels for BWRs and in future for PHWRs. The plant can also manufacture mixed oxide fuel for prototype fast breeder reactor (PFBR) and development work in this regard has already started. The paper describes the MOX fuel manufacturing technology and quality control techniques presently in use at the plant. The irradiation experience of the lead MOX assemblies in BWRs is also briefly discussed. The key areas of interest for future developments in MOX fuel fabrication technology and Pu utilisation are identified. (author)

Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

International Nuclear Information System (INIS)

Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

2013-01-01

For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

MOX-fuel inherent proliferation-protection due to {sup 231}Pa admixture

Energy Technology Data Exchange (ETDEWEB)

Kryuchkov, E.F.; Glebov, V.B.; Apse, V.A.; Shmelev, A.N. [Moscow Engineering Physics Institute (State University), Moscow (Russian Federation)

2003-07-01

The proliferation protection levels of MOX-fuel containing small additions of protactinium are evaluated for equilibrium closed and open cycles of a light-water reactor (LWR).Analysis of the ways to the proliferation protection of MOX-fuel by small {sup 231}Pa addition and comparison of this way with another options for giving MOX-fuel the proliferation self-protection property enable us to make the 3 following conclusions: -1) Unique nature of protactinium as a small addition to MOX-fuel is determined by the following properties: - Protactinium is available in the nature (uranium ore) as a long-lived mono-isotope {sup 231}Pa, - under neutron irradiation, {sup 231}Pa is converted into {sup 232}U, which is a long-term source of high energy gamma-radiation and practically non-separable from main fuel mass, - essentially, {sup 231}Pa is a high-quality burnable neutron absorber. -2) From the proliferation self-protection point of view, nuclear fuel cycle closure with fuel recycle is a preferable option because, under this condition, introduction of protactinium into MOX-fuel allows to create the inherent radiation barrier which is in action during full cycle of fuel management at the level corresponding to the accepted today criterion of the Spent Fuel Standard (SFS). In particular, the considered example of multiple MOX-fuel recycle with small addition of {sup 231}Pa (0.2% HM) at each cycle demonstrates a possibility to reach the proliferation protection level of fresh MOX-fuel corresponding to once irradiated fuel with the same cooling time. In this case, the lethal dose (at 30 cm distance from fuel assembly) is received within the minute range. -3) Introduction of {sup 231}Pa into MOX-fuel composition in amount of 0.5% HM allows to prolong action of the SFS from 100 to 200 years. If {sup 231}Pa content is increased up to 5% HM, then MOX-fuel conserves the proliferation self-protection property in respect to short-term unauthorized actions for 200-year period of its

The status of BNFL's MOX project

International Nuclear Information System (INIS)

Edwars, John; Cooch, Julian P.; Slater, Michel W.

2002-01-01

Full text: In the late 1980s BNFL decided to enter the MOX fuel fabrication business to support our reprocessing business and return the plutonium product to our customers in the useable form of MOX fuel. The first phase of the strategy was to gain some irradiation experience for MOX produced by our own Short Binderless Route (SBR) process. To achieve this the MOX Demonstration Facility (MDF) was built at Sellafield and 28 MOX fuel assemblies were produced up to 1998 that were loaded into PWRs in Europe. In 1994, BNFL started the construction of their large scale MOX production plant, SMP. The design and construction of the plant and supporting facilities was completed some years ago and the commissioning of the plant with uranium commenced around June 1999. In October 2001, the UK Government provided BNFL with the approval to operate SMP with plutonium. On 20 December 2001, the UK Regulators gave BNFL their approval to start plutonium operations. This paper summarises the approach used to commission SMP and describes some of the lessons learnt during the commissioning phase of the project and the start up of the plant with plutonium. An explanation of our experience obtaining a licence to operate the plant is provided together with a description of the changes we have made to ensure that the quality of the product from SMP can be guaranteed. Finally, the paper summarises the experience BNFL has gained during irradiating MOX fuel produced by the SBR process and explains how the data compares with that available for UO2 and supports the in reactor use of MOX fuel made in SMP. (author)

Performance of MOX fuel: An overview of the experimental programme of the OECD Halden Reactor Project and review of selected results

International Nuclear Information System (INIS)

Wiesenack, W.; McGrath, M.

2000-01-01

The OECD Halden Reactor Project has defined an extensive experimental programme related to MOX fuels which is being executed with the objective to provide a performance data base similar to that available for UO 2 . In addition to utilising fresh MOX fuel and re-instrumented segments from LWR irradiations to high burnup, the concept of inert matrix fuel is being addressed. The irradiation in the Halden reactor is performed in rigs allowing steady state, power ramping and cyclic operation. In-pile data are obtained from instrumentation such as fuel centreline thermocouples, pressure transducers, fuel and cladding elongation detectors, and movable gauges for measuring the diametral deformation. Various phenomena can be assessed in this way, e.g. thermal performance, swelling and densification, PCMI and fission gas release. The paper describes the objectives of various experiments and provides examples of temperature, pressure and cladding elongation measurements performed on MOX fuel. Salient results are related to the threshold for the onset of significant fission gas release and the relaxation behaviour in a power ramp-PCMI situation. (author)

Overall models and experimental database for UO2 and MOX fuel increasing performance

International Nuclear Information System (INIS)

Bernard, L.C.; Blanpain, P.

2001-01-01

COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

MOX fuel fabrication at AECL

International Nuclear Information System (INIS)

Dimayuga, F.C.; Jeffs, A.T.

1995-01-01

Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

Probability of Criticality for MOX SNF

International Nuclear Information System (INIS)

P. Gottlieb

1999-01-01

The purpose of this calculation is to provide a conservative (upper bound) estimate of the probability of criticality for mixed oxide (MOX) spent nuclear fuel (SNF) of the Westinghouse pressurized water reactor (PWR) design that has been proposed for use. with the Plutonium Disposition Program (Ref. 1, p. 2). This calculation uses a Monte Carlo technique similar to that used for ordinary commercial SNF (Ref. 2, Sections 2 and 5.2). Several scenarios, covering a range of parameters, are evaluated for criticality. Parameters specifying the loss of fission products and iron oxide from the waste package are particularly important. This calculation is associated with disposal of MOX SNF

A utility analysis of MOX recycling policy

International Nuclear Information System (INIS)

Pfaeffli, J.L.

1990-01-01

The author presents the advantages of recycling of plutonium and uranium from spent reactor fuel assemblies as follows: natural uranium and enrichment savings, mixed oxide fuel (MOX) fuel assembly cost, MOX compatibility with plant operation, high burnups, spent MOX reprocessing, and non-proliferation aspects.Disadvantages of the recycling effort are noted as well: plutonium degradation with time, plutonium availability, in-core fuel management, administrative authorizations by the licensings authorities, US prior consent, and MOX fuel fabrication capacity. Putting the advantages and disadvantages in perspective, it is concluded that the recycling of MOX in light water reactors represents, under the current circumstances, the most appropriate way of making use of the available plutonium

Recycling of MOX fuel for LWRs

International Nuclear Information System (INIS)

Joo, Hyung Kook; Oh, Soo Youl

1992-01-01

The status and issues related to the thermal recycling of reprocessed nuclear fuels have been reviewed. It is focused on the use of reprecessed plutonium in the form of mixed oxide (MOX) for a light water reactor and the review on reprocessing and fabrication processes is beyond the scope. In spite of the difference in the nuclear characteristics between plutonium and uranium isotopes, the neutronics behavior in a core with MOX fuels is similar to that with normal uranium fuels. However, since the neutron spectrum is hardened in a core with MOX, the Doppler, viod, and moderator temperature coefficients become more negative and the control rod and boron worths are slightly reduced. Therefore, the safety will be evaluated carefully in addition to the core neutronics analysis. The MOX fuel rod behavior related to the rod performance such as the pellet to clad interaction and fission gas release is also similar to that of uranium rods, and no specific problem arises. Substituting MOX fuels for a portion of uranium fuels, it is estimated that the savings be about 25% in uranium ore and 10% in uranium enrichment service requirements. The use of MOX fuel in LWRs has been commercialized in European countries including Germany, France, Belgium, etc., and a demonstration program has been pursued in Japan for the commercial utilization in the late 1990s. Such a worldwide trend indicates that the utilization of MOX fuel in LWRs is a proven technology and meets economics criteria. (Author)

Top-MOX fuel solution: strategies, challenges, opportunities

International Nuclear Information System (INIS)

Breitenstein, P.; Vo Van, V.

2014-01-01

TOP-MOX is a nuclear fuel solution and product developed by AREVA and successfully implemented in Europe. It allows utilities burning plutonium (instead of enriched uranium) even when this plutonium is not stemming from own reprocessed used fuel - that is third party plutonium. The important challenges for utilities along with TOP-MOX implementation are legal/patrimonial Pu-ownership issues and general economical aspects. Available sponsorship of such plutonium permits UO2 competitive market prices. For new MOX customers licensing and technical aspects come along. Further AREVA proposes a flexible solution which is called 'TOP-MOX pre-cycling'. This involves making available third party plutonium for fuel fabrication and reactor use pending the utilities' final strategic fuel cycle decision. The paper gives insight into and analyses the impacts of allowing customers the implementation of a TOP-MOX program with focus on Pu-ownership, economics, technical and legal aspects as well as the impact on used MOX management and final waste management. (authors)

High moderation MOX cores for effective use of plutonium in LWRs

International Nuclear Information System (INIS)

Hamamoto, Kazuko; Kanagawa, Takashi; Hiraiwa, Koji; Sakurada, Koichi; Moriwaki, Masanao; Aoyama, Motoo; Yamamoto, Toru; Ueji, Masao

2001-01-01

Conceptual design studies have been performed for high moderation full MOX cores aiming at increasing fissile Pu consumption rate (ratio of the consumed to the loaded fissile Pu) and reducing residual Pu in discharged MOX fuel. The BWR cores studied have hydrogen to heavy metal ratio(H/HM) of 5.9 with increasing water rods and 7.0 with reducing a fuel rod diameter based on a reference 9x9 fuel (H/HM=4.9) of ABWR. The PWR cores studied have H/HM of 5.0 and 6.0 with reducing a fuel rod diameter based on a reference 17x17 fuel (H/HM=4.0) of APWR. Equilibrium core design and plant safety analyses showed that those high moderation cores have compatibility with ABWR and APWR. The fissile Pu consumption rate is 22% larger than the full MOX cores with reference fuel of ABWR and 50% for APWR. The core performance and compatibility has been also evaluated in the condition of multi-recycle of Pu in these high moderation cores. Study has been conducted to evaluate the effect of introducing these high moderation cores in the fuel cycle of Japan. It shows that the high moderation cores produce 26% more cumulative electricity and reduce 22% stock of the fissile Pu by 2050 than the reference cores. (author)

ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor

International Nuclear Information System (INIS)

Garnier, J.C.; Ikegami, T.

1993-01-01

Description of program or function: In order to intercompare the characteristics of the different reactors considered for Pu recycling, in terms of neutron economy, minor actinide production, uranium content versus Pu burning, the NSC Working Party on Physics of Plutonium Recycling (WPPR) is setting up several benchmark studies. They cover in particular the case of the evolution of the Pu quality and Pu fissile content for Pu recycling in PWRs; the void coefficient in PWRs partly fuelled with MOX versus Pu content; the physics characteristics of non-standard fast reactors with breeding ratios around 0.5. The following benchmarks are considered here: - Fast reactors: Pu Burner MOX fuel, Pu Burner metal fuel; - PWRs: MOX recycling (bad quality Pu), Multiple MOX recycling

Models for MOX fuel behaviour. A selective review

International Nuclear Information System (INIS)

Massih, Ali R.

2006-01-01

This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO 2 fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO 2 . In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO 2 fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO 2 fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO 2 vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO 2 . This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation

Models for MOX fuel behaviour. A selective review

Energy Technology Data Exchange (ETDEWEB)

Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

2006-12-15

This report reviews the basic physical properties of light water reactor mixed-oxide (MOX) fuel comprising nuclear characteristics, thermal properties such as melting temperature, thermal conductivity, thermal expansion, and heat capacity, and compares these with properties of conventional UO{sub 2} fuel. These properties are generally well understood for MOX fuel and are well described by appropriate models developed for engineering analysis. Moreover, certain modelling approaches of MOX fuel in-reactor behaviour, regarding densification, swelling, fission product gas release, helium release, fuel creep and grain growth, are evaluated and compared with the models for UO{sub 2}. In MOX fuel the presence of plutonium rich agglomerates adds to the complexity of fuel behaviour on the micro scale. In addition, we survey the recent fuel performance experience and post irradiation examinations on several types of MOX fuel types. We discuss the data from these examinations, regarding densification, swelling, fission product gas release and the evolution of the microstructure during irradiation. The results of our review indicate that in general MOX fuel has a higher fission gas release and helium release than UO{sub 2} fuel. Part of this increase is due to the higher operating temperatures of MOX fuel relative to UO{sub 2} fuel due to the lower thermal conductivity of MOX material. But this effect by itself seems to be insufficient to make for the difference in the observed fission gas release of UO{sub 2} vs. MOX fuel. Furthermore, the irradiation induced creep rate of MOX fuel is higher than that of UO{sub 2}. This effect can reduce the pellet-clad interaction intensity in fuel rods. Finally, we suggest that certain physical based approaches discussed in the report are implemented in the fuel performance code to account for the behaviour of MOX fuel during irradiation.

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)

Late effects following inhalation of mixed oxide (U,PuO2) mox aerosol in the rat

International Nuclear Information System (INIS)

Griffiths, N.; Van Der Meeren, A.; Fritsch, P.; Maximilien, R.

2008-01-01

Exposure to alpha-emitting particles is a potential long-term health risk to workers in nuclear fuel fabrication plants. Mixed Oxide (MOX: U,PuO 2 ) fuels containing low percentages of plutonium obtained from spent nuclear fuels are increasingly employed and in the case of accidental contamination by inhalation or wounds may result in the development of late-occurring pathologies such as lung cancer. However the long term risks particularly with regard to lung cancer are to date unclear. In the case of MOX the risk may indeed be different from that assigned to the individual components, plutonium and uranium. Several factors are influential (i) the dissolution of Pu depends on the physico-chemical properties, for example risk of lung cancer is increased 10 fold after Pu(NO 3 ) 2 as compared with PuO 2 . (ii) The solubility of Pu is variable whether delivered as PuO 2 or contained within MOX. (iii) The risk of cancer appears to increase with spatial homogeneity of the lung alpha dose. The objective of this study was to investigate the long term effects in rat lungs following MOX aerosol inhalation of similar particle size containing 2.5 or 7.1% Pu. Conscious rats were exposed to MOX aerosols using a 'nose-only' system and kept for their entire life (2-3 years). Different Initial Lung Deposits (ILDs) were obtained using different concentrations of the MOX suspension. Lung total alpha activity was determined in vivo at intervals over the study period by external counting as well as at autopsy in order to estimate the total lung dose. Anatomo-pathological and immunohistochemical analyses were performed on fixed lung tissue after euthanasia. The frequencies of lung pathologies and tumours were determined on lung sections at several different levels. In addition, autoradiography of lung sections was performed in order to assess the spatial localisation of a activity. Inhalation of MOX at ILD ranging from 1-20 kBq resulted in lung pathologies (90% of exposed rats

Synthesis and biological activity of novel series of 4-methoxy, and 4,9-dimethoxy-5-substituted furo[2,3-g]-1,2,3-benzoxathiazine-7,7-dioxide derivatives.

Science.gov (United States)

El-Sawy, Eslam R; Ebaid, Manal S; Abo-Salem, Heba M; El-Hallouty, Salwa; Kassem, Emad M; Mandour, Adel H

2014-05-01

A novel series of 4-methoxy, and 4,9-dimethoxy-5-substituted furo[2,3-g]-1,2,3-benzoxathiazine-7,7-dioxide derivatives 3a,b, 10a-g and 11a-g were prepared in good yields via the reaction of 4-methoxy (1a) and 4,7-dimethoxy-5-acetyl-6-hydroxybenzofurans (1b) and their α,β-unsaturated keto derivatives 6a-g and 7a-g with chlorosulfonyl isocyanate (CSI). On the other hand, N-chlorosulfonyl carbamate derivatives 4a,b, 12a,b and 13a,b were prepared and allowed to react with piperidine to give the corresponding N-piperidinosulfonyl carbamate derivatives 5a,b, 14a,b and 15a,b, respectively. Sixteen new target compounds 3a,b, 10a-g, and 11a-g were tested for their DPPH radical-scavenging, and in vitro antiproliferative activity against A-549, MCF7 and HCT-116 cancer cell lines. Compounds 10a, 11c, 11e, and 11g showed moderate DPPH radical-scavenging activity compared to ascorbic acid at 100 μg/mL. 4,9-Dimethoxy-5-substituted styrylfuro[3,2-g]-1,2,3-benzoxathiazine-7,7-dioxides 11a, 11b, and 11c were found to be highly active against A-549 and HCT-116 cancer cell lines with IC50 values ranging from 0.02 to 0.08 μmol/mL compared to doxorubicin with IC50 = 0.04 and 0.06 μmol/mL, respectively.

Public acceptance of MOX - fuel

International Nuclear Information System (INIS)

Huettmann, A.; Reddehase, C.G.

1995-01-01

In the Federal Republic of Germany 'Plutonium-Business' got fresh nutrient because of the carried out licensing of the use of Mixed Oxide (MOX)-fuel LWR and in connection with the negative attitude of the Hessian authorities, who are responsible for the licensing procedures of the production of MOX-fuel in the Siemens-factories at Hanau. The opponents of the peaceful use of nuclear energy try with the emotive expression 'Plutonium' (Pu) a frontal attack against the use of nuclear energy in Germany. They justify their actions with so-called safety deficits of the plants and increased danger of cancer in case of using MOX-fuel. (orig./HP)

Degradation of Acid Orange 7 by peroxymonosulfate activated with the recyclable nanocomposites of g-C3N4 modified magnetic carbon.

Science.gov (United States)

Guo, Furong; Lu, Jiahua; Liu, Qing; Zhang, Ping; Zhang, Aiqing; Cai, Yingjie; Wang, Qiang

2018-08-01

Carbon-based catalysts have attracted high attention since they are greener and cheaper, while magnetic nanomaterials are very useful in environmental application because of the easy recovery and operation given by the magnetic separability. Therefore, graphitic carbon nitride modified magnetic carbon nanocomposites Fe 3 O 4 @C/g-C 3 N 4 was prepared herein for the first time as a new carbon-based catalyst for the activation of peroxymonosulfate (PMS). The catalytic properties of Fe 3 O 4 @C/g-C 3 N 4 in activating PMS for the degradation of Acid Orange 7 (AO 7), a model organic pollutant, were investigated. AO 7 degradation efficiency was significantly enhanced after modification of Fe 3 O 4 @C with g-C 3 N 4 , and the composite Fe 3 O 4 @C/g-C 3 N 4 from loading of 5 wt% g-C 3 N 4 and calcined at 300 °C for 30 min exhibited the best performance. AO 7 could be efficiently decolorized using the "Fe 3 O 4 @C/C 3 N 4 (5%) + PSM" system within the pH range of 2-6, and 97% of AO 7 could be removed in 20 min without pH adjustment (pH = 4). Radical quenching and EPR studies confirmed that both sulfate and hydroxyl radicals produced from PMS activation were the active species responsible for the oxidation of AO 7. The degradation mechanism was suggested based on the experimental results and XPS analyses. It was proposed that the CO groups on the carbon surface of Fe 3 O 4 @C rather than the CO in g-C 3 N 4 played a key role as the active sites for PMS activation. The catalyst was magnetically separable and displayed good stability and reusability, thus providing a potentially green catalyst for sustainable remediation of organic pollutants. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

Code Analyses Supporting PIE of Weapons-Grade MOX Fuel

International Nuclear Information System (INIS)

Ott, Larry J.; Bevard, Bruce Balkcom; Spellman, Donald J.; McCoy, Kevin

2010-01-01

The U.S. Department of energy has decided to dispose of a portion of the nation's surplus weapons-grade plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating the fuel in commercial power reactors. Four lead test assemblies (LTAs) were manufactured with weapons-grade mixed oxide (WG-MOX) fuel and irradiated in the Catawba Nuclear Station Unit 1, to a maximum fuel rod burnup of ∼47.3 GWd/MTHM. As part of the fuel qualification process, five rods with varying burnups and initial plutonium contents were selected from one assembly and shipped to the Oak Ridge National Laboratory (ORNL) for hot cell examination. ORNL has provided analytical support for the post-irradiation examination (PIE) of these rods via extensive fuel performance modeling which has aided in instrument settings and PIE data interpretation. The results of these fuel performance simulations are compared in this paper with available PIE data.

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

Safety evaluation on MOX new fuel at marine transport

International Nuclear Information System (INIS)

Tsumune, Daisuke; Ito, Chihiro; Saegusa, Toshiari; Maruyama, Koki

2000-01-01

In the Central Research Institute of Electric Power Industry, in order to confirm effects of MOX new fuel on the public are as small as possible even when its marine transport goes down, some exposed radiation dose has previously conducted on imaginary shipwreck of marine transport on used nuclear fuel, plutonium dioxide, and high level return glass solid. Under a base of such informations, some investigations on safety on marine transport of the MOX new fuel was conducted. On September, 1999, five transport vessels of the MOX new fuel was at first transported on marine. The value of five times of estimated exposed radiation dose (max. 8.1 x 10 -8 mSv/y) corresponds to an evaluation result assumed by shipwreck in marine transport this time. As a result, it was found that the exposed radiation dose estimated on this case would be sufficiently less than an effective dose equivalent limit (1 mSv/y) of public exposure according to the recommendation of ICRP in both coastal and oceanic areas. (G.K.)

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Energy Technology Data Exchange (ETDEWEB)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-02-26

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

2001-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

Transport of MOX fuel from Europe to Japan

International Nuclear Information System (INIS)

2002-01-01

The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

International Nuclear Information System (INIS)

Chang, G.S.

2004-01-01

In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

Energy Technology Data Exchange (ETDEWEB)

Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

2000-01-01

This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

Energy Technology Data Exchange (ETDEWEB)

Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

1997-08-01

Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

Impact of the thermal scattering law of H in H_2O on the isothermal temperatures reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

International Nuclear Information System (INIS)

Scotta, J.P.; Noguere, G.; Bernard, D.; Santamarina, A.; Damian, J.I.M.

2016-01-01

The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France). The interpretation of the core residual reactivity measured between 6 to 80 C. degrees (by step of 5 C. degrees) was performed with the Monte-Carlo code TRIPOLI-4"R. The nuclear data from the JEFF-3.1.1 library were used in the calculations. 3 different thermal scattering laws of hydrogen in light water were tested in order to evaluate their impact on the MISTRAL calculations. The thermal scattering laws of interest were firstly those recommended in JEFF-3.1.1 and ENDF/BVII.1 and also that recently produced at the atomic center of Bariloche (CAB, Argentina) with molecular dynamic simulations. The present work indicates that the calculation-to-experimental bias is (0.4 ± 0.3) pcm/C. degree in the UOX core and (1.0 ± 0.3) pcm/C. degree in the MOX cores, when the JEFF-3.1.1 library is used. An improvement is observed over the whole temperature range with the CAB model. The calculation-to-experimental bias vanishes for the UOX core (0.02 pcm/C. degree) and becomes close to 0.7 pcm/C. degree for the MOX cores. The magnitude of these bias have to be connected to the typical value of the temperature reactivity coefficient that ranges from 5 pcm/C. degree at Beginning Of Cycle (BOC) up to 50 pcm/C. degrees at End Of Cycle (EOC), in PWR conditions. (authors)

MOX fuel fabrication, in reactor performance and improvement

International Nuclear Information System (INIS)

Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

1998-01-01

In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

R7-binding protein targets the G protein β5/R7-regulator of G protein signaling complex to lipid rafts in neuronal cells and brain

Directory of Open Access Journals (Sweden)

Zhang Jian-Hua

2007-09-01

Full Text Available Abstract Background Heterotrimeric guanine nucleotide-binding regulatory proteins (G proteins, composed of Gα, Gβ, and Gγ subunits, are positioned at the inner face of the plasma membrane and relay signals from activated G protein-coupled cell surface receptors to various signaling pathways. Gβ5 is the most structurally divergent Gβ isoform and forms tight heterodimers with regulator of G protein signalling (RGS proteins of the R7 subfamily (R7-RGS. The subcellular localization of Gβ 5/R7-RGS protein complexes is regulated by the palmitoylation status of the associated R7-binding protein (R7BP, a recently discovered SNARE-like protein. We investigate here whether R7BP controls the targeting of Gβ5/R7-RGS complexes to lipid rafts, cholesterol-rich membrane microdomains where conventional heterotrimeric G proteins and some effector proteins are concentrated in neurons and brain. Results We show that endogenous Gβ5/R7-RGS/R7BP protein complexes are present in native neuron-like PC12 cells and that a fraction is targeted to low-density, detergent-resistant membrane lipid rafts. The buoyant density of endogenous raft-associated Gβ5/R7-RGS protein complexes in PC12 cells was similar to that of lipid rafts containing the palmitoylated marker proteins PSD-95 and LAT, but distinct from that of the membrane microdomain where flotillin was localized. Overexpression of wild-type R7BP, but not its palmitoylation-deficient mutant, greatly enriched the fraction of endogenous Gβ5/R7-RGS protein complexes in the lipid rafts. In HEK-293 cells the palmitoylation status of R7BP also regulated the lipid raft targeting of co-expressed Gβ5/R7-RGS/R7BP proteins. A fraction of endogenous Gβ5/R7-RGS/R7BP complexes was also present in lipid rafts in mouse brain. Conclusion A fraction of Gβ5/R7-RGS/R7BP protein complexes is targeted to low-density, detergent-resistant membrane lipid rafts in PC12 cells and brain. In cultured cells, the palmitoylation status of

MOX fuel fabrication: Technical and industrial developments

International Nuclear Information System (INIS)

Lebastard, G.; Bairiot, H.

1990-01-01

The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

MOX fuel irradiation behaviour: Results from X-ray microbeam analysis

International Nuclear Information System (INIS)

Walker, C.T.; Goll, W.; Matsumura, T.

1997-01-01

The behaviour of plutonium, xenon and caesium were investigated in two sections of irradiated MOX fuel produced by the OCOM process. In one fuel (OCOM30), the MOX agglomerates contained 18 wt% fissile plutonium, and had a low volume fraction of 0.17; in the other (OCOM15) the agglomerates contained 9 wt% fissile plutonium, and had a high volume fraction of 0.34. Both fuels had been irradiated under normal power reactor conditions to a burn-up of approximately 44 GWd/t. The main aim of the work was to establish whether the above differences in composition affected the percentage fission gas released by the fuels. Since U/Pu interdiffusion did not occurred during the irradiation, both fuels remained inhomogeneous on the microscopic scale. However, the concentration of plutonium in the MOX agglomerates decreases by about 50% as a result of fission, whereas the plutonium content of the UO 2 matrix increased by about a factor of four to approximately 2 wt% due to neutron capture by 238 U. The agglomerates in the OCOM15 fuel generally exhibited a finer structure due to the lower burn-up. More than 80% of the fission gas had been released from the oxide lattice of the MOX agglomerates in both fuels. However, a very high fraction of this gas precipitated and remained in the pore structure of the agglomerates. Consequently, puncturing revealed that for both fuels the percentage of gas released to the rod free volume increased from less than 0.5% at 10 GWd/t to a maximum of 3.5% at 45 GWd/t. The conclusion is that the percentage of gas released by MOX fuel is largely unaffected of the level of inhomogeneity of the fuel. In both fuels caesium showed near complete retention in both the MOX agglomerates and the UO 2 matrix. (author). 8 refs, 11 figs, 3 tabs

Progress in researches on MOX fuel pellet producing technology in China

International Nuclear Information System (INIS)

Hu Xiaodan

2010-01-01

Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

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)

Overview of MOX fuel fabrication achievements

International Nuclear Information System (INIS)

Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.

2000-01-01

Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)

MOX recycling-an industrial reality

International Nuclear Information System (INIS)

Shallo, G.D.F.

1996-01-01

Reprocessing and plutonium recycling have now attained industrial maturity in France and Europe. Specifically, mixed-oxide (MOX) fuel is fabricated and used in light water reactors (LWRs) in satisfactory operating conditions. The utilities and the fuel cycle industry experience no technical difficulties, and European recycling programs are growing steadily, from 18 reactors in operation today up to 50 expected around the year 2000, putting the system reprocessing-recycling in coherence: 25 t of plutonium will then be used each year to produce the electricity equivalence of 25 millions tons of oil. Plutonium recycling in MOX fuel in current LWRs proves to be technically safe and economically competitive and meets natural resource savings and environmental protection objectives. And recycling responds properly to the nonproliferation concerns. Such an industrial experience gives a unique reference for weapons plutonium disposition through MOX use in reactors

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

1996-01-01

Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

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.

MOX fuel fabrication and utilisation in LWRs worldwide

International Nuclear Information System (INIS)

Provost, J.-L.; Schrader, M.; Nomura, S.

2000-01-01

Early in the development of the nuclear programme, a large part of the countries using nuclear energy has studied the reprocessing and recycling option in order to develop a safe conditioning of fission products and to recycle fissile materials in reactors. In the sixties, the feasibility of recycling plutonium in LWRs has been successfully demonstrated by several experimentations of MOX rod irradiations in different countries. Based on the background of the MOX behaviour collected during the seventies and on the results of the important MOX experimentation program implemented during this period, a large part of the European utilities decided at the beginning of the eighties to use MOX fuel in LWRs on an industrial scale. The main goals of the utilities were to use as a fuel an available fissile material and to control the stockpile of separated plutonium. Today, the understanding of the behaviour of plutonium fuel has grown significantly since the launch of the first R and D programmes on LWR and FR MOX fuels. Plutonium oxide physical and neutron behaviour is well known, its modelling is now available as well as experimentally validated. Up to now, more than 750 tHM MOX fuel (more than 2000 FAs) have been loaded in 29 PWRs and in 2 BWRs in Europe, corresponding to the recycling of about 35 t of plutonium. Reprocessing/recycling technology has reached maturity in the main nuclear industry countries. Spent fuel reprocessing and recycling of the separated fissile materials remains the main option for the back-end cycle. Today, the operation of MOX-recycling LWRs is considered satisfactory. Experience feedback shows that, in global terms, MOX cores behaviour is equivalent to that of UO 2 cores in terms of operation and safety. (author)

Criticality safety philosophy for the Sellafield MOX plant

International Nuclear Information System (INIS)

Edge, Jane; Gulliford, Jim

2003-01-01

The Sellafield MOX Plant (SMP) has been operational since 2001, blending plutonium dioxide from THORP reprocessing operations, with uranium dioxide to produce Mixed Oxide (MOX) fuel elements. In handling the quantities of fuel associated with a commercial fuel fabrication plant, it is necessary to impose criticality controls. Plutonium dioxide (PuO 2 ), uranium dioxide (UO 2 ) and recycled MOX are mixed together in batches. An Engineered Protection System (EPS) prevents the production of MOX powder in excess of 20w/o Pu(fissile)/(Pu+U), achieved through the combination of a weight-based' system and a diverse 'neutron monitoring' radiometric system. The 'neutron monitoring' component of the EPS determines the fissile enrichment of the batch of MOX powder, based on pessimistic isotopic requirements of the PuO 2 feedstock powder. Guaranteeing the maximum MOX enrichment of 20w/o Pu(fissile)/(Pu + U) at an early stage of the fuel manufacturing process enables the criticality safety assessor to demonstrate that normal operations are deterministically safe. This paper describes in detail the EPS at the front end of plant and the engineered and operational protection in downstream areas. In addition plant operational experience in producing the first fuel assemblies is discussed. (author)

Mox pellet reference material

International Nuclear Information System (INIS)

Perolat, J.P.

1991-01-01

A first batch of MOX pellets certified in plutonium and uranium has been prepared and characterised in France to meet the needs of laboratories which are engaged upon destructive analysis for safeguards purposes especially in fuel fabrication plants. The pellets sintering has been obtained in a special fabrication to achieve an homogeneity better than 0.1%. The plutonium and uranium characterisation by chemical analysis has been carried out by two laboratories using at least two different methods. 1 fig., 5 refs

Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

Energy Technology Data Exchange (ETDEWEB)

Carbajo, J.J.

2005-05-27

The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

Radial power density distribution of MOX fuel rods in the HBWR

International Nuclear Information System (INIS)

Koo, Yang Hyun; Joo, Hyung Kook; Lee, Byung Ho; Sohn, Dong Seong

1999-07-01

Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3 Catalysts for CO Oxidation

Directory of Open Access Journals (Sweden)

Hongmei Qin

2015-04-01

Full Text Available Conventional supported Pt catalysts have often been prepared by loading Pt onto commercial supports, such as SiO2, TiO2, Al2O3, and carbon. These catalysts usually have simple metal-support (i.e., Pt-SiO2 interfaces. To tune the catalytic performance of supported Pt catalysts, it is desirable to modify the metal-support interfaces by incorporating an oxide additive into the catalyst formula. Here we prepared three series of metal oxide-modified Pt catalysts (i.e., Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3, where M = Al, Fe, Co, Cu, Zn, Ba, La for CO oxidation. Among them, Pt/CoOx/SiO2, Pt/CoOx/TiO2, and Pt/CoOx/Al2O3 showed the highest catalytic activities. Relevant samples were characterized by N2 adsorption-desorption, X-ray diffraction (XRD, transmission electron microscopy (TEM, H2 temperature-programmed reduction (H2-TPR, X-ray photoelectron spectroscopy (XPS, CO temperature-programmed desorption (CO-TPD, O2 temperature-programmed desorption (O2-TPD, and CO2 temperature-programmed desorption (CO2-TPD.

New Digital Metal-Oxide (MOx Sensor Platform

Directory of Open Access Journals (Sweden)

Daniel Rüffer

2018-03-01

Full Text Available The application of metal oxide gas sensors in Internet of Things (IoT devices and mobile platforms like wearables and mobile phones offers new opportunities for sensing applications. Metal-oxide (MOx sensors are promising candidates for such applications, thanks to the scientific progresses achieved in recent years. For the widespread application of MOx sensors, viable commercial offerings are required. In this publication, the authors show that with the new Sensirion Gas Platform (SGP a milestone in the commercial application of MOx technology has been reached. The architecture of the new platform and its performance in selected applications are presented.

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

Energy Technology Data Exchange (ETDEWEB)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

2000-07-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

International Nuclear Information System (INIS)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

2000-01-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Experimental microstructures MOX fuels elaboration

International Nuclear Information System (INIS)

Gotta, M.J.; Dubois, S.; Lechelle, J.; Sornay, P.

2000-01-01

In order to propose a new MOX fuel, owning higher combustion rate, studies are realized at the CEA in collaboration with Cogema, EDF and Framatome. New microstructures of MOX are looked for around two approaches: the grains size and the plutonium distribution. These approaches are presented and discussed in this paper. The first one develops big grains microstructures obtained, either with anionic (sulfur), or cationic (Cr 2 O 3 ) additives. The second one concerns the CER-CER type composite microstructures. (A.L.B.)

Dissolution behavior of PFBR MOX fuel in nitric acid

International Nuclear Information System (INIS)

Kelkar, Anoop; Kapoor, Y.S.; Singh, Mamta; Meena, D.L.; Pandey, Ashish; Bhatt, R.B.; Behere, P.G.

2017-01-01

Present paper describes the dissolution characteristics of PFBR MOX fuel (U,Pu)O 2 in nitric acid. An overview of batch dissolution experiments, studying the percentage dissolution of uranium and plutonium in (U, Pu)O 2 MOX sintered pellets with different percentage of PuO 2 with reference to time and nitric acid concentration are described. 90% of uranium and plutonium of PFBR MOX gets dissolves in 2 hrs and amount of residue increases with the decrease in nitric acid concentration. Overall variation in percentage residue in PFBR MOX fuel after dissolution test also described. (author)

Structure-function analysis of STING activation by c[G(2',5')pA(3',5')p] and targeting by antiviral DMXAA.

Science.gov (United States)

Gao, Pu; Ascano, Manuel; Zillinger, Thomas; Wang, Weiyi; Dai, Peihong; Serganov, Artem A; Gaffney, Barbara L; Shuman, Stewart; Jones, Roger A; Deng, Liang; Hartmann, Gunther; Barchet, Winfried; Tuschl, Thomas; Patel, Dinshaw J

2013-08-15

Binding of dsDNA by cyclic GMP-AMP (cGAMP) synthase (cGAS) triggers formation of the metazoan second messenger c[G(2',5')pA(3',5')p], which binds the signaling protein STING with subsequent activation of the interferon (IFN) pathway. We show that human hSTING(H232) adopts a "closed" conformation upon binding c[G(2',5')pA(3',5')p] and its linkage isomer c[G(2',5')pA(2',5')p], as does mouse mSting(R231) on binding c[G(2',5')pA(3',5')p], c[G(3',5')pA(3',5')p] and the antiviral agent DMXAA, leading to similar "closed" conformations. Comparing hSTING to mSting, 2',5'-linkage-containing cGAMP isomers were more specific triggers of the IFN pathway compared to the all-3',5'-linkage isomer. Guided by structural information, we identified a unique point mutation (S162A) placed within the cyclic-dinucleotide-binding site of hSTING that rendered it sensitive to the otherwise mouse-specific drug DMXAA, a conclusion validated by binding studies. Our structural and functional analysis highlights the unexpected versatility of STING in the recognition of natural and synthetic ligands within a small-molecule pocket created by the dimerization of STING. Copyright © 2013 Elsevier Inc. All rights reserved.

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

MOXE: An X-ray all-sky monitor for Soviet Spectrum-X-Gamma Mission

Science.gov (United States)

Priedhorsky, W.; Fenimore, E. E.; Moss, C. E.; Kelley, R. L.; Holt, S. S.

1989-01-01

A Monitoring Monitoring X-Ray Equipment (MOXE) is being developed for the Soviet Spectrum-X-Gamma Mission. MOXE is an X-ray all-sky monitor based on array of pinhole cameras, to be provided via a collaboration between Goddard Space Flight Center and Los Alamos National Laboratory. The objectives are to alert other observers on Spectrum-X-Gamma and other platforms of interesting transient activity, and to synoptically monitor the X-ray sky and study long-term changes in X-ray binaries. MOXE will be sensitive to sources as faint as 2 milliCrab (5 sigma) in 1 day, and cover the 2 to 20 KeV band.

Development of MOX facilities and the impact on the nuclear fuel markets

International Nuclear Information System (INIS)

Patterson, J.

1990-01-01

Mixed-oxide (MOX) fuel is nearing maturity as a fuel supply option. This paper briefly reviews the history and current status of the MOX fuel market, including the projected increase in demand for MOX fuel as more plutonium becomes available from the operation of commercial irradiated fuel reprocessing plants in Europe. The uncertainties of such projected demand are discussed, together with the anticipated requirements from the next generation of MOX fabrication plants. The impact of the growing demand for MOX fuel is assessed in the traditional sectors of the uranium fuel cycle. Finally, the author turns to a generalized treatment of the economic aspects of MOX fuel utilization, showing the financially attractive regimes of MOX use which will benefit nuclear power utilities and continue to ensure that MOX fuel can consolidate its position as a mature fuel supply option in those countries that have opted to recycle their spent fuel

Experience from start-ups of the first ANITA Mox plants.

Science.gov (United States)

Christensson, M; Ekström, S; Andersson Chan, A; Le Vaillant, E; Lemaire, R

2013-01-01

ANITA™ Mox is a new one-stage deammonification Moving-Bed Biofilm Reactor (MBBR) developed for partial nitrification to nitrite and autotrophic N-removal from N-rich effluents. This deammonification process offers many advantages such as dramatically reduced oxygen requirements, no chemical oxygen demand requirement, lower sludge production, no pre-treatment or requirement of chemicals and thereby being an energy and cost efficient nitrogen removal process. An innovative seeding strategy, the 'BioFarm concept', has been developed in order to decrease the start-up time of new ANITA Mox installations. New ANITA Mox installations are started with typically 3-15% of the added carriers being from the 'BioFarm', with already established anammox biofilm, the rest being new carriers. The first ANITA Mox plant, started up in 2010 at Sjölunda wastewater treatment plant (WWTP) in Malmö, Sweden, proved this seeding concept, reaching an ammonium removal rate of 1.2 kgN/m³ d and approximately 90% ammonia removal within 4 months from start-up. This first ANITA Mox plant is also the BioFarm used for forthcoming installations. Typical features of this first installation were low energy consumption, 1.5 kW/NH4-N-removed, low N₂O emissions, started up at Sundets WWTP in Växjö, Sweden, reached full capacity with more than 90% ammonia removal within 2 months from start-up. By applying a nitrogen loading strategy to the reactor that matches the capacity of the seeding carriers, more than 80% nitrogen removal could be obtained throughout the start-up period.

Analysis of canine herpesvirus gB, gC and gD expressed by a recombinant vaccinia virus.

Science.gov (United States)

Xuan, X; Kojima, A; Murata, T; Mikami, T; Otsuka, H

1997-01-01

The genes encoding the canine herpesvirus (CHV) glycoprotein B (gB), gC and gD homologues have been reported already. However, products of these genes have not been identified yet. Previously, we have identified three CHV glycoproteins, gp 145/112, gp80 and gp47 using a panel of monoclonal antibodies (MAbs). To determine which CHV glycoprotein corresponds to gB, gC or gD, the putative genes of gB, gC, and gD of CHV were inserted into the thymidine kinase gene of vaccinia virus LC16mO strain under the control of the early-late promoter for the vaccinia virus 7.5-kilodalton polypeptide. We demonstrated here that gp145/112, gp80 and gp47 were the translation products of the CHV gB, gC and gD genes, respectively. The antigenic authenticity of recombinant gB, gC and gD were confirmed by a panel of MAbs specific for each glycoprotein produced in CHV-infected cells. Immunization of mice with these recombinants produced high titers of neutralizing antibodies against CHV. These results suggest that recombinant vaccinia viruses expressing CHV gB, gC and gD may be useful to develop a vaccine to control CHV infection.

Autoradiographic studies on the distribution of 14C-5,7-dihydroxytryptamine in the brain of new-born rat

International Nuclear Information System (INIS)

Lappe, U.

1982-01-01

The distribution of intracisternally injected 14 C-5,7-dihydroxy tryptamine (5,7-DHT) in the central nervous system of new-born rat is studied by means of autoradiography. The radio-active neurotoxin is incorporated into the neurones of all known serotonine nucleus groups. This labelling allows a detailed demonstration of the topography of the serotonine neurones in the brain stem of the new-born rat and to compare it with systems obtained by other methods. Serotonine neurones were mapped in 22 representative frontal sections through the brain stem. 14 C-5,7-DHT is incorporated into noradrenergic neurones, too. However, labelling is less marked than in serotonergic neurones. 14 C-5,7-DHT is incorporated at small quantities into the following extraneural elements: fibroblasts of the pia mater/arachnoidea, some endothelical cells of pial vessels, epithelial cells of the plexus choroideus, and subependymal macrophages. 6 h after injection of 25 μg 14 C-5,7-DHT, the vast majority of serotonergic neurones reveal strong degenerative changes which are irreversible. (orig./MG) [de

Studies of Flexible MOX/LEU Fuel Cycles

International Nuclear Information System (INIS)

Adams, M.L.; Alonso-Vargas, G.

1999-01-01

This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report

MX 8: the next generation high capacity system for the transport of fresh MOX fuel

International Nuclear Information System (INIS)

Potelle, F.; Issard, H.

1998-01-01

The choice of reprocessing policy was made a long time ago in France, leading to the development of an advanced Pu recycling industry. In 1987, Saint Laurent was the first French reactor to be loaded with fresh MOX fuel. Transnucleaire, then in charge of transport packaging development, created the FS 69 concept, derived from the classical RCC concept for the transport of UO 2 fresh fuel. On the other hand, Cogema, as the main actor in the field of fuel cycle and thus in transport matters, developed the associated security truck and security caisson in order to provide the transport system with the acceptable Physical Protection devices required by French Authorities. As a whole, the security truck and the FS 69 have now been used for more than ten years with a remarkable level of efficiency and safety. Indeed, more than 600 fresh MOX fuel elements have been delivered, without any incident, both regarding safety or fuel integrity requirements. But, as a matter of fact, the replacement of FS 69 transport system is now scheduled for several reasons. First of all, the burnups achieved with UO 2 fuel progressed together with its enrichment within the last ten years, and the MOX 'equivalence' also implies that its Pu content be increased to enhance its reactor performances: from 5.25 % of Pu content today, the MOX fuel will reach 7% tomorrow, and almost 10% the day after tomorrow. Lastly, the reprocessing/recycling policy has been confirmed and amplified, leading to an increasing number of 'moxified' reactors. As a consequence, the French utility (EDF), the fuel designer (Fragema, the joint venture between Framatome and Cogema), the fuel manufacturer (Cogema), and the transporter (Transnucleaire) joined in a specific working group devoted to the development of the MX 8, the next generation high capacity system for the land transport of MOX fuel. (authors)

Molecular and crystal structure of nido-9-C5H5N-11-I-7,8-C2B9H10: supramolecular architecture via hydrogen bonding X-H...I (X = B, C)

International Nuclear Information System (INIS)

Polyanskaya, T.M.

2006-01-01

A monocrystal X-ray diffraction study of a new iodine-containing cluster compound 9-(pyridine)-11-iodo-decahydro-7,8-dicarba-nido-undecaborane [9-C 5 H 5 N-11-I-7,8-C 2 B 9 H 10 ] has been performed. Crystal data: C 7 H 15 B 9 NI, M = 337.39, monoclinic, space group P2 1 /c, unit cell parameters: a=9.348(1) A, b=11.159(1) A, c=13.442(2) A, β=98.13(1) deg, V=1388.1(5) A 3 , Z=4, d calc = 1.614 g/cm 3 , T = 295 K, F(000)=648, μ=2.276 mm -1 . The structure was solved by a direct method and refined in the full-matrix anisotropic approximation (isotropic for hydrogen atoms) to final agreement factors R 1 = 0.0254, wR 2 = 0.0454 for 2437 I hkl >2σ I from 3590 measured I hkl (an Enraf-Nonius CAD-4 diffractometer, λMoK α , graphite monochromator, θ/2θ-scanning). The molecules are joined into a supramolecular assembly by hydrogen bonds X-H...I (X = B, C) [ru

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

A NOVEL APPROACH TO FIND OPTIMIZED NEUTRON ENERGY GROUP STRUCTURE IN MOX THERMAL LATTICES USING SWARM INTELLIGENCE

Directory of Open Access Journals (Sweden)

M. AKBARI

2013-12-01

Full Text Available Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2–PUO2 (MOX is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2–PuO2 (MOX lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

RIA tests in CABRI with MOX fuel

International Nuclear Information System (INIS)

Schmitz, F.; Papin, J.; Gonnier, C.

2000-01-01

Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

Determination of thorium and plutonium in AHWR experimental (Th, 1%Pu)O2 MOX fuel after microwave dissolution

International Nuclear Information System (INIS)

Fulzele, Ajit K.; Malav, R.K.; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Das, D.K.; Prakash, Amrit; Behere, P.G.; Afzal, Mohd

2013-01-01

This paper describes determination of thorium and plutonium in experimental (Th, 1%Pu)O 2 AHWR (Advanced Heavy Water Reactor) MOX fuel samples after dissolution by microwave. Time taken to dissolve ∼ 2g of MOX sample by conventional IR heating technique in conc. HNO 3 + 0.05 M HF mixture is about 35-40 hours while using microwave dissolution technique it is ∼ 2 hours. Hence, with the help of microwave dissolution technique analysis time for each sample has been reduced from week to a day. The PuO 2 content (wt%) in the MOX pellets was within specification limit, (1.0±0.1)%. (author)

BNFL assessment of methods of attaining high burnup MOX fuel

International Nuclear Information System (INIS)

Brown, C.; Hesketh, K.W.; Palmer, I.D.

1998-01-01

It is clear that in order to maintain competitiveness with UO 2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO 2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

Analysis of ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}D{sub 2}){gamma}{sub 5g}, {gamma}{sub 3g} and ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}L{sub 6}){gamma}{sub 1g}, a{gamma}{sub 5g} two-photon absorption spectra of Cs{sub 2}NaYF{sub 6}:Eu{sup 3+}

Energy Technology Data Exchange (ETDEWEB)

Ning Lixin; Wang Dianyuan; Xia Shangda [Structure Research Laboratory, Academica Sinica, Department of Physics, University of Science and Technology of China, Heifei, Anhui (China); Thorne, Jonathan R.G. [Inorganic Chemistry Laboratory, Department of Chemistry, University of Oxford (United Kingdom); Tanner, Peter A. [Department of Biology and Chemistry, City University of Hong Kong, Kowloon (China)

2002-04-15

The direct calculation of transition line strengths and relative intensities is presented for two intraconfigurational two-photon absorption (TPA) transitions of Eu{sup 3+} in the cubic Cs{sub 2}NaYF{sub 6} host. Crystal field wavefunctions were utilized for the initial and final f{sup N}-electron states and various approaches were used in constructing all the 4f{sup N-1} 5d{sup 1} intermediate-state wavefunctions. The calculated relative intensities of the ({sup 7}F{sub 0}) {gamma}{sub 1g}{yields}({sup 5}D{sub 2}){gamma}{sub 5g}, {gamma}{sub 3g} TPA transitions are in reasonable agreement with experiment. The neglect of J-mixing in the initial state has only a small effect upon the calculation, whereas the neglect of spin-orbit couplings within the initial and terminal states drastically reduces the calculated transition linestrengths, but does not markedly change the intensity ratios. In the case of the ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}L{sub 6}){gamma}{sub 1g}, a{gamma}{sub 5g} transitions, serious discrepancies between experiment and theory are found if the intermediate states are constructed from a 4f{sup 5} core comprising free ion states and the 5d{sup 1} crystal field states. Satisfactory agreement is, however, found when the 4f{sup 5} crystal field states are utilized in constructing the intermediate states. The contributions to the transition moment have been evaluated for various Hamiltonian terms and the results are discussed. (author)

A MOX fuel attribute monitor

International Nuclear Information System (INIS)

Bliss, Mary; Jordan, David V.; Barnett, Debra S.; Redding, Rebecca L.; Pearce, Stephen K.

2007-01-01

Euratom performs safeguards monitoring of Fresh MOX fuel for domestic power production in the European Union. Video cameras monitor the reactor storage ponds. If video surveillance is lost for a certain amount of time a measurement is required to verify that no fuel was diverted. The attribute measurement to verify the continued presence of MOX fuel is neutron emission. Ideally this measurement would be made without moving or handling the fuel rod assembly. A prototype attribute measurement system was made using scintillating neutron sensitive glass waveguides developed by Pacific Northwest National Laboratory. Short lengths (5-20 cm) of the neutron sensitive fiber were mechanically spliced to 15 m lengths of commercial high numerical aperture fiber optic cable (Ceramoptec Optran Ultra 0.44). The light detector is a Hamamatsu R7400P photomultiplier tube. An electronics package was built to use the sensors with a GBS Elektronik MCA-166 multichannel analyzer and user interface. The MCA-166 is the system most commonly used by Euratom inspectors. It can also be run from a laptop computer using Maestro (Ortec) or other software. A MCNP model was made to compare to measurements made with several neutron sources including NIST traceable 252 Cf

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)

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

International Nuclear Information System (INIS)

Mertyurek, Ugur; Gauld, Ian C.

2016-01-01

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

Validation of MOX fuel through recent BELGONUCLEAIRE international programmes

International Nuclear Information System (INIS)

Basselier, J.; Maldague, T.; Lippens, M.

1997-01-01

The paper reviews the present experience of BELGONUCLEAIRE in promoting and managing international programmes dedicated to improvement and updating of MOX fuel data bases on what concerns core physics and rod behaviour with a view of assist all MOX fuel designers and users in their validation and modelization work. All these programmes were completed or will be completed with the support of numerous international organizations deeply concerned by MOX recycling strategies. (author). 9 figs, 2 tabs

Kinetics for exchange of imino protons in the d(C-G-C-G-A-A-T-T-C-G-C-G) double helix and in two similar helices that contain a G . T base pair, d(C-G-T-G-A-A-T-T-C-G-C-G), and an extra adenine, d(C-G-C-A-G-A-A-T-T-C-G-C-G).

Science.gov (United States)

Pardi, A; Morden, K M; Patel, D J; Tinoco, I

1982-12-07

The relaxation lifetimes of imino protons from individual base pairs were measured in (I) a perfect helix, d(C-G-C-G-A-A-T-T-C-G-C-G), (II) this helix with a G . C base pair replaced with a G . T base pair, d(C-G-T-G-A-A-T-T-C-G-C-G), and (III) the perfect helix with an extra adenine base in a mismatch, d(C-G-C-A-G-A-A-T-T-C-G-C-G). The lifetimes were measured by saturation recovery proton nuclear magnetic resonance experiments performed on the imino protons of these duplexes. The measured lifetimes of the imino protons were shown to correspond to chemical exchange lifetimes at higher temperatures and spin-lattice relaxation times at lower temperatures. Comparison of the lifetimes in these duplexes showed that the destabilizing effect of the G . T base pair in II affected the opening rate of only the nearest-neighbor base pairs. For helix III, the extra adenine affected the opening rates of all the base pairs in the helix and thus was a larger perturbation for opening of the base pairs than the G . T base pair. The temperature dependence of the exchange rates of the imino proton in the perfect helix gives values of 14-15 kcal/mol for activation energies of A . T imino protons. These relaxation rates were shown to correspond to exchange involving individual base pair opening in this helix, which means that one base-paired imino proton can exchange independent of the others. For the other two helices that contain perturbations, much larger activation energies for exchange of the imino protons were found, indicating that a cooperative transition involving exchange of at least several base pairs was the exchange mechanism of the imino protons. The effects of a perturbation in a helix on the exchange rates and the mechanisms for exchange of imino protons from oligonucleotide helices are discussed.

Foundations for the definition of MOX fuel quality requirements

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vanderborck, Y.

1991-01-01

The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, the fuel enrichment process takes place in the manufacturing plant, the radioactivity of Pu imposes handling constraints, Pu ages quite rapidly, altering its isotopic composition during storage, the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart form UO 2 fuel. (orig.)

Bi-Modal Model for Neutron Emissions from PuO2 and MOX Holdup

International Nuclear Information System (INIS)

Menlove, Howard; Lafleur, Adrienne

2015-01-01

O 2 particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm 3 , but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O 2 in the PuO 2 as well as any impurity contamination such as H 2 O. To obtain a more quantitative estimate of the neutron (�

gamma-ray spectra measurements for long cooled MOX spent fuels

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Kobayashi, Iwao

1993-09-01

Gamma-ray spectra of spent fuels have important informations in the estimation of burnup rate, concentration of fission products, cooling time and etc. which are required in the fuel loading control of reactors and special nuclear materials accountancy from the view point of safe guard. Although, some available data are given about uranium dioxide fuels, few data are given about uranium and plutonium dioxide mixtures (MOX fuels). Especially, there is few data about MOX fuels which are irradiated in thermal reactors and cooled more than ten years. Gamma-ray spectra are measured for PuO 2 -UO 2 fuel rods (IFA-159, IFA-160) which are irradiated at HBWR in Norway up to 9,420 and 5,340MWd/t respectively. Gamma-ray spectra had been measured about the two fuels ten years ago at the spent fuel pond of Japan Demonstration Reactor (JPDR). The objectives of this measurement is to know how decayed the gamma-ray spectra in these ten years and some fission products are there which are effective to estimate burnup rate of spent MOX fuels. (author)

Analysis of high moderation full MOX BWR core physics experiments BASALA

International Nuclear Information System (INIS)

Ishii, Kazuya; Ando, Yoshihira; Takada, Naoyuki; Kan, Taro; Sasagawa, Masaru; Kikuchi, Tsukasa; Yamamoto, Toru; Kanda, Ryoji; Umano, Takuya

2005-01-01

Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO 2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO 2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores. (author)

Plasmid-mediated AmpC-type beta-lactamase isolated from Klebsiella pneumoniae confers resistance to broad-spectrum beta-lactams, including moxalactam.

Science.gov (United States)

Horii, T; Arakawa, Y; Ohta, M; Ichiyama, S; Wacharotayankun, R; Kato, N

1993-01-01

Klebsiella pneumoniae NU2936 was isolated from a patient and was found to produce a plasmid-encoded beta-lactamase (MOX-1) which conferred resistance to broad spectrum beta-lactams, including moxalactam, flomoxef, ceftizoxime, cefotaxime, and ceftazidime. Resistance could be transferred from K. pneumoniae NU2936 to Escherichia coli CSH2 by conjugation with a transfer frequency of 5 x 10(-7). The structural gene of MOX-1 (blaMOX-1) was cloned and expressed in E. coli HB101. The MIC of moxalactam for E. coli HB101 producing MOX-1 was > 512 micrograms/ml. The apparent molecular mass and pI of this enzyme were calculated to be 38 kDa and 8.9, respectively. Hg2+ and Cu2+ failed to block enzyme activity, and the presence of EDTA in the reaction buffer did not reduce the enzyme activity. However, clavulanate and cloxacillin, serine beta-lactamase inhibitors, inhibited the enzyme activity competitively (Kis = 5.60 and 0.35 microM, respectively). The kinetic study of MOX-1 suggested that it effectively hydrolyzed broad-spectrum beta-lactams. A hybridization study confirmed that blaMOX-1 is encoded on a large resident plasmid (pRMOX1; 180 kb) of strain NU2936. By deletion analysis, the functional region was localized within a 1.2-kb region of the plasmid. By amino acid sequencing, 18 of 33 amino acid residues at the N terminus of MOX-1 were found to be identical to those of Pseudomonas aeruginosa AmpC. These findings suggest that MOX-1 is a plasmid-mediated AmpC-type beta-lactamase that provides enteric bacteria resistance to broad-spectrum beta-lactams, including moxalactam. Images PMID:8517725

An overview of economic and technical issues related to LWR MOX fuel usage

International Nuclear Information System (INIS)

Malone, J.P.; Varley, G.; Goldstein, L.

1999-01-01

This paper will present comparisons of the economics of MOX versus UO 2 fuels. In addition to the economics of the front end, the scope of the comparison will include the back end of the fuel cycle. Management of spent MOX fuel assemblies presents utilities with some technical issues that can complicate spent fuel pool operation. Alternative spent fuel management methods, such as dry storage of spent MOX fuel assemblies, will also be discussed. Differences in decay heat loads versus time for spent MOX and UO 2 fuel assemblies will be presented. This difference is one of the main problems confronting spent fuel managers relative to MOX. The difference in decay heat loads will serve as the basis for a performance overview of the various spent fuel technologies available today. The economics of the front end of MOX will be presented relative to UO 2 fuel. Availability of MOX manufacturing capability will also be discussed, along with a discussion of its impact on future MOX fabrication prices. The in-core performance of MOX will be compared to that of UO 2 fuel with similar performance characteristics. The information will include highlights of nuclear design and related operational considerations such as: Reactivity reduction with burnup is slower for MOX fuel than for UO 2 fuel; Spectral hardening resulting in lower control rod worths and a lower soluble boron worth; and more negative moderator, void and fuel temperature coefficients. A comparison of Westinghouse and ABB-CE core designs for use on disposition of weapons MOX in 12- and 18-month cycles will be presented. (author)

MOX fuel cycle technologies for medium and long term deployment. Proceedings

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

More than thirty years of reactor experience using MOX fuel as well as the fabrication of 2000 MOX assemblies with the use of 85 t of Pu separated from spent fuel from power reactors indicates that the recycling of plutonium as MOX fuel in LWRs has become a mature industry. The number of countries engaged in plutonium recycling could be increasing in the near future, aiming for the reduction of stockpiles of separated plutonium from earlier and existing reprocessing contracts. Economic and strategic considerations are the main factors on which to base such a decision to use MOX. Transport of MOX fuel assemblies is a vital element in these recycle programmes but could have the potential to be a weak link in the chain. To avoid problems, it is essential that sufficient numbers of transport flasks of the required types, licensed for the increasing Pu contents, be made available in a timely manner to keep pace with the planned increases in fabrication rates. Despite the excellent safety records for radioactive and MOX transports over many decades, continuous attention should be drawn to establishing the transport modalities, buffer stores, secure vehicles, and transport routes, at the same time accounting for public sensitivities on radioactive transports in general and MOX transport in particular. A large number of technical presentations updated and reconfirmed the good and almost defect-free performance of MOX fuel at increasingly high burn-up levels. MOX fuel is designed to meet the same operational and safety criteria as uranium fuels under equivalent conditions. This is also confirmed by the parallel development of design codes to accommodate the special characteristics of MOX. Integral and specific parameter testing of MOX fuel in normal and off-normal operation is under way in a number of countries with particular emphasis on high burnup behaviour. Here the important contributions of the OECD/NEA Halden BWR programme should be mentioned. The reactor

MOX fuel cycle technologies for medium and long term deployment. Proceedings

International Nuclear Information System (INIS)

2000-01-01

More than thirty years of reactor experience using MOX fuel as well as the fabrication of 2000 MOX assemblies with the use of 85 t of Pu separated from spent fuel from power reactors indicates that the recycling of plutonium as MOX fuel in LWRs has become a mature industry. The number of countries engaged in plutonium recycling could be increasing in the near future, aiming for the reduction of stockpiles of separated plutonium from earlier and existing reprocessing contracts. Economic and strategic considerations are the main factors on which to base such a decision to use MOX. Transport of MOX fuel assemblies is a vital element in these recycle programmes but could have the potential to be a weak link in the chain. To avoid problems, it is essential that sufficient numbers of transport flasks of the required types, licensed for the increasing Pu contents, be made available in a timely manner to keep pace with the planned increases in fabrication rates. Despite the excellent safety records for radioactive and MOX transports over many decades, continuous attention should be drawn to establishing the transport modalities, buffer stores, secure vehicles, and transport routes, at the same time accounting for public sensitivities on radioactive transports in general and MOX transport in particular. A large number of technical presentations updated and reconfirmed the good and almost defect-free performance of MOX fuel at increasingly high burn-up levels. MOX fuel is designed to meet the same operational and safety criteria as uranium fuels under equivalent conditions. This is also confirmed by the parallel development of design codes to accommodate the special characteristics of MOX. Integral and specific parameter testing of MOX fuel in normal and off-normal operation is under way in a number of countries with particular emphasis on high burnup behaviour. Here the important contributions of the OECD/NEA Halden BWR programme should be mentioned. The reactor

77 FR 50519 - Agency Information Collection Activities: Forms G-325, G-325A, G-325B, and G-325C; Extension of a...

Science.gov (United States)

2012-08-21

... DEPARTMENT OF HOMELAND SECURITY U.S. Citizenship and Immigration Services [OMB Control Number 1615... Department of Homeland Security (DHS), U.S. Citizenship and Immigration Services (USCIS) will be submitting... collection: Forms G-325, G-325A, G-325B, and G-325C; U.S. Citizenship and Immigration Services (USCIS). (4...

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)

New metal-organic frameworks of [M(C6H5O7)(C6H6O7)(C6H7O7)(H2O)] . H2O (M=La, Ce) and [Ce2(C2O4)(C6H6O7)2] . 4H2O

International Nuclear Information System (INIS)

Weng Shengfeng; Wang, Yun-Hsin; Lee, Chi-Shen

2012-01-01

Two novel materials, [M(C 6 H 5 O 7 )(C 6 H 6 O 7 )(C 6 H 7 O 7 )(H 2 O)] . H 2 O (M=La(1a), Ce(1b)) and [Ce 2 (C 2 O 4 )(C 6 H 6 O 7 ) 2 ] . 4H 2 O (2), with a metal-organic framework (MOF) were prepared with hydrothermal reactions and characterized with photoluminescence, magnetic susceptibility, thermogravimetric analysis and X-ray powder diffraction in situ. The crystal structures were determined by single-crystal X-ray diffraction. Compound 1 crystallized in triclinic space group P1-bar (No. 2); compound 2 crystallized in monoclinic space group P2 1 /c (No. 14). The structure of 1 is built from a 1D MOF, composed of deprotonated citric ligands of three kinds. Compound 2 contains a 2D MOF structure consisting of citrate and oxalate ligands; the oxalate ligand arose from the decomposition in situ of citric acid in the presence of Cu II ions. Photoluminescence spectra of compounds 1b and 2 revealed transitions between the 5d 1 excited state and two levels of the 4f 1 ground state ( 2 F 5/2 and 2 F 7/2 ). Compounds 1b and 2 containing Ce III ion exhibit a paramagnetic property with weak antiferromagnetic interactions between the two adjacent magnetic centers. - Graphical Abstract: [M(C 6 H 5 O 7 )(C 6 H 6 O 7 )(C 6 H 7 O 7 )(H 2 O)] . H 2 O (M=La(1a), Ce(1b)) and [Ce 2 (C 2 O 4 )(C 6 H 6 O 7 ) 2 ] . 4H 2 O (2)—with 1D and 2D structures were synthesized and characterized. Highlights: ► Two MOF – [M(C 6 H 5 O 7 )(C 6 H 6 O 7 )(C 6 H 7 O 7 )(H 2 O)] . H 2 O (M=La(1a), Ce(1b)) and [Ce 2 (C 2 O 4 )(C 6 H 6 O 7 ) 2 ] . 4H 2 O (2) – with 1D and 2D structures. ► The adjacent chains of the 1D framework were correlated with each other through an oxalate ligand to form a 2D layer structure. ► The source of the oxalate ligand was the decomposition in situ of citric acid oxidized in the presence of Cu II ions.

Memento. Maritime transport of MOX fuels from Europe to Japan

International Nuclear Information System (INIS)

1999-07-01

The maritime transport of MOX fuels from Europe to Japan represents the last of the 3 steps of transport of the nuclear fuel reprocessing-recycling program settled between ORC (Japan), BNFL (UK) and Cogema (France). This document summarizes the different aspects of this program: the companies concerned, the physical protection measures, the US-Japan agreements (accompanying warship), the in-depth safety, the handling of MOX fuels (containers and ships), and the Japan MOX fuel needs. (J.S.)

Modelling the actual behaviour of the MOX fuel by a micromechanical analysis in non-uniform transformation fields

International Nuclear Information System (INIS)

Largenton, R.

2012-01-01

This research thesis aimed at developing a model based on scale change to assess more precisely the distribution of local thermo-mechanical fields within a heterogeneous medium as MOX fuel. The analysis method is a non-uniform transformation field analysis (NTFA) which is adapted to the problem of scale change in presence of a coupling between dissipative and elastic effects. More precisely, the author addressed the development of a NTFA model based on specific three-phase and three-dimensional microstructures which are typical of the MOX fuel in an in-service operation. The first part proposes an overview of knowledge and use of MOX. It recalls the context and the industrial problematic associated with this fuel: operating principles for a 900 MWe PWR, fuel fabrication processes, fuel morphologies and structural and microstructural consequences. It addresses local mechanisms within each phase during irradiation, and presents the approach methodology regarding scale change. The second part reports the representation and analysis in complete fields of multiphase particle-based composites (MOX type) in order to determine the representative elementary volume and the local behaviour of each phase. The third part reports the extension of the NTFA approach to 3D aspects, free deformations, ageing and optimization. The last part compares the NTFA approach with the incremental two-phase and three-phase Mori-Tanaka models

Validation study of core analysis methods for full MOX BWR

International Nuclear Information System (INIS)

2013-01-01

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

Validation study of core analysis methods for full MOX BWR

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

JENDL-3.2 performance in analyses of MISTRAL critical experiments for high-moderation MOX cores

International Nuclear Information System (INIS)

Takada, Naoyuki; Hibi, Koki; Ishii, Kazuya; Ando, Yoshihira; Yamamoto, Toru; Ueji, Masao; Iwata, Yutaka

2001-01-01

NUPEC and CEA have launched an extensive experimental program called MISTRAL to study highly moderated MOX cores for the advanced LWRs. The analyses using SRAC system and MVP code with JENDL-3.2 library are in progress on the experiments of the MISTRAL and the former EPICURE programs. Various comparisons have been made between calculation results and measurement values. (author)

The need for integral critical experiments with low-moderated MOX fuels

International Nuclear Information System (INIS)

2004-01-01

The use of MOX fuel in commercial reactors is a means of burning plutonium originating from either surplus weapons or reprocessed irradiated uranium fuel. This requires the fabrication of MOX assemblies on an industrial scale. The OECD/NEA Expert Group on Experimental Needs for Criticality Safety has highlighted MOX fuel manufacturing, as an area in which there is a specific need for additional experimental data for validation purposes. Indeed, integral experiments with low-moderated MOX fuel are either scarce or not sufficiently accurate to provide an appropriate degree of validation of nuclear data and computer codes. New and accurate experimental data would enable a better optimisation of the fabrication process by decreasing the uncertainties in the determination of multiplication factors of configurations such as the homogenization of MOX powders. In this context, the OECD/NEA Nuclear Science Committee organised a workshop to address the following topics: expression and justification of the need for critical or near-critical experiments employing low-moderated MOX fuels; proposals for experimental programmes to address these needs; prospects for an international co-operative programme. The workshop was held at OECD headquarters in Paris on 14-15 April 2004. (author)

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)

Recycling schemes of Americium targets in PWR/MOX cores

International Nuclear Information System (INIS)

Maldague, Th.; Pilate, S.; Renard, A.; Harislur, A.; Mouney, H.; Rome, M.

1999-01-01

From the orientation studies performed so far, both ways to recycle Am in PWR/MOX cores, homogeneous in MOX or heterogeneous in target pins, appear feasible, provided that enriched UO 2 is used as support of the MOX fuel. Multiple recycling can then proceed and stabilize Pu and Am quantities. With respect to the Pu multiple recycling strategy, recycling Am in addition needs 1/3 more 235 U, and creates 3 times more Curium. Thus, although feasible, such a fuel cycle is complicated and brings about a significant cost penalty, not quantified yet. The advantage of the heterogeneous option is to allow to manage in different ways the Pu in MOX fuel and the Am in target pins. For example, should Am remain combined to Cm after reprocessing, the recycling of a mix of Am+Cm could be deferred to let Cm transform into Pu before irradiation. The Am+Cm targets could also stay longer in the reactor, so as to avoid further reprocessing if possible. (author)

Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

International Nuclear Information System (INIS)

Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

1999-09-01

Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantly affects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs

Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

1999-09-01

Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantlyaffects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs.

WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

International Nuclear Information System (INIS)

J.W. Davis

1996-01-01

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so

A human repair gene ERCC5 is involved in group G xeroderma pigmentosum

International Nuclear Information System (INIS)

Shiomi, Tadahiro

1994-01-01

In E. coli, ultraviolet-induced DNA damage is removed by the coordinated action of UVR A, B, C, and D proteins (1). In Saccharomyces cerevisiae, more than ten genes have been reported to be involved in excision repair (2). The nucleotide excision repair pathway has been extensively studied in these organisms. To facilitate studying nucleotide excision repair in mammalian cells. Ultraviolet-sensitive rodent cell mutants have been isolated and classified into 11 complementation groups (9,10). The human nucleotide excision repair genes which complement the defects of the mutants have been designated as the ERCC (excision repair cross-complementing) genes; a number is added to refer to the particular rodent complementation group that is corrected by the gene. Recently, several human DNA repair genes have been cloned using rodent cell lines sensitive to ultraviolet. These include ERCC2 (3), ERCC3 (4), and ERCC6 (5), which correspond to the defective genes in the ultraviolet-sensitive human disorders xeroderma pigmentosum (XP) group D (6) and group B (4), and Cockayne's syndrome (CS) group B (7), respectively. The human excision repair gene ERCC5 was cloned after DNA-mediated gene transfer of human HeLa cell genomic DNA into the ultraviolet-sensitive mouse mutant XL216, a member of rodent complementation group 5 (11,12) and the gene was mapped on human chromosome 13q32.3-q33.1 by the replication R-banding fluorescence in situ hybridization method (13). The ERCC5 cDNA encodes a predicted 133 kDa nuclear protein that shares some homology with product of the yeast DNA repair gene RAD 2. Transfection with mouse ERCC5 cDNA restored normal levels of ultraviolet-resistance to XL216 cells. Microinjection of ERCC5 cDNA specifically restored the defect of XP group G cells (XP-G) as measured by unscheduled DNA synthesis (UDS), and XP-G cells stably transformed with ERCC5 cDNA showed nearly normal ultraviolet resistance. (J.P.N.)

Quantitative Ethylene Measurements with MOx Chemiresistive Sensors at Different Relative Air Humidities

Directory of Open Access Journals (Sweden)

Matic Krivec

2015-11-01

Full Text Available The sensitivity of two commercial metal oxide (MOx sensors to ethylene is tested at different relative humidities. One sensor (MiCS-5914 is based on tungsten oxide, the other (MQ-3 on tin oxide. Both sensors were found to be sensitive to ethylene concentrations down to 10 ppm. Both sensors have significant response times; however, the tungsten sensor is the faster one. Sensor models are developed that predict the concentration of ethylene given the sensor output and the relative humidity. The MQ-3 sensor model achieves an accuracy of ±9.2 ppm and the MiCS-5914 sensor model predicts concentration to ±7.0 ppm. Both sensors are more accurate for concentrations below 50 ppm, achieving ±6.7 ppm (MQ-3 and 5.7 ppm (MiCS-5914.

The somatic FAH C.1061C>A change counteracts the frequent FAH c.1062+5G>A mutation and permits U1snRNA-based splicing correction.

Science.gov (United States)

Scalet, Daniela; Sacchetto, Claudia; Bernardi, Francesco; Pinotti, Mirko; van de Graaf, Stan F J; Balestra, Dario

2018-05-01

In tyrosinaemia type 1(HT1), a mosaic pattern of fumarylacetoacetase (FAH) immunopositive or immunonegative nodules in liver tissue has been reported in many patients. This aspect is generally explained by a spontaneous reversion of the mutation into a normal genotype. In one HT1 patient carrying the frequent FAH c.1062+5G>A mutation, a second somatic change (c.1061C>A) has been reported in the same allele, and found in immunopositive nodules. Here, we demonstrated that the c.1062+5G>A prevents usage of the exon 12 5' splice site (ss), even when forced by an engineered U1snRNA specifically designed on the FAH 5'ss to strengthen its recognition. Noticeably the new somatic c.1061C>A change, in linkage with the c.1062+5G>A mutation, partially rescues the defective 5'ss and is associated to trace level (~5%) of correct transcripts. Interestingly, this combined genetic condition strongly favored the rescue by the engineered U1snRNA, with correct transcripts reaching up to 60%. Altogether, these findings elucidate the molecular basis of HT1 caused by the frequent FAH c.1062+5G>A mutation, and demonstrate the compensatory effect of the c.1061C>A change in promoting exon definition, thus unraveling a rare mechanism leading to FAH immune-reactive mosaicism.

Parametric study on co-precipitation of U/Th in MOX fuel of AHWR

International Nuclear Information System (INIS)

Tiwari, S.K.; Swaroopa Lakshmi, Y.; Nath, Baidurjya; Setty, D.S.; Kalyana Krishnan, G.; Saibaba, N.

2015-01-01

During manufacturing of Mixed Oxide Fuel (MOX) pellets for Advance Heavy Water Reactor (AHWR-LEU), around 30% rejected MOX pellets are generated in every cycle. These rejected MOX pellets are dissolved in nitric acid for recovery of U/Th. The recovered U/Th is recycled for production of MOX pellets. MOX pellets of varying compositions are used in AHWR fuel. Dissolution of MOX pellets in nitric acid is a challenging task because of its low surface area and longer dissolution times. High normal nitric acid is used in order to increase rate of dissolution, which in turn results in generation of high free acidity solution which influences the precipitation characteristics of Uranium (VI) by oxalic acid. Oxalic acid precipitation helps in generation of nitric acid which can be used for dissolution there by effectively facilitating nil effluent generation. Precipitation by oxalic acid unlike ammonia has advantage of zero liquid effluent discharge by complete recycle of oxalate filtrate to dissolution section. In the present work, the effect of various parameters like free acidity, residence time, concentration of oxalic acid, initial concentration of uranium and thorium etc. on the precipitation of U(VI) and Th(IV) in nitrate media by oxalic acid was carried out. The precipitated powder was subjected to various morphological evaluations like particle size etc. Study of various parameters on the co-precipitation of uranium and thorium by oxalic acid was carried out. It was observed that complete precipitation (> 99.9%) of thorium as oxalate does not depend on free acidity range (1- 6 N). Excess oxalic acid is not required for complete precipitation of thorium oxalate. The precipitation of uranyl oxalate varies with initial free acidity of solution. Uranyl oxalate precipitation does not take place at and above 5 N of free acidity

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

International Nuclear Information System (INIS)

Fridman, E.; Kliem, S.

2011-01-01

Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

The rates of G:C[yields]T:A and G:C[yields]C:G transversions at CpG dinucleotides in the human factor IX gene

Energy Technology Data Exchange (ETDEWEB)

Ketterling, R.P.; Vielhaber, E.; Sommer, S.S. (Mayo Clinic/Foundation, Rochester, MN (United States))

1994-05-01

The authors have identified eight independent transversions at CpG in 290 consecutive families with hemophilia B. These eight transversions account for 16.3% of all independent transversions in the sample, yet the expected frequency of CpG transversions at random in the factor IX gene is only 2.6% (P<0.1). The aggregate data suggest that the two types of CpG transversions (G:C[yields]T:A and G:C[yields]C:G) possess similar mutation rates (24.8 [times] 10[sup [minus]10] and 20.6 [times] 10[sup [minus]10], respectively), which are about fivefold greater than the comparable rates for transversions at non-CpG dinucleotides. The enhancement of transversions at CpG suggest that the model by which mutations occur at CpG may need to be reevaluated. The relationship, if any, between deamination of 5-methyl cytosine and enhancement of transversions at CpG remains to be defined. 28 refs., 2 tabs.

Evaluation of full MOX core capability for a 900 MWe PWR

International Nuclear Information System (INIS)

Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

1996-01-01

Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

Glove box adaptation, installation and commissioning of WD-XRF system for determination of PuO2 in MOX fuel samples

International Nuclear Information System (INIS)

Aher, Sachin; Pandey, Ashish; Khan, F.A.; Das, D.K.; Kumar, Surendra; Behere, P.G.; Mohd Afzal

2015-01-01

safe plutonium-handling criterion. Leak rate achieved during testing is 0.05% of Glove Box volume which classifies these Glove Boxes as a Class-I containment as per ISO Standard and also meeting the ASTM Standard C-852 recommended value for plutonium handling glove boxes. This paper highlights the technique for adaptation of process equipments inside the Glove Box, necessary additional features and changes in system and its components for GB adaptations. Details of Glove Box installations, selection and assembly of Glove Box Panels, Ventilation, Provisions of various service lines (e.g. Gas Line, Vacuum Line, Electrical Connections) and commissioning criterion of Glove Box for plutonium handling is discussed in this paper. (author)

Advanced high throughput MOX fuel fabrication technology and sustainable development

International Nuclear Information System (INIS)

Krellmann, Juergen

2005-01-01

The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In

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

International Nuclear Information System (INIS)

O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

1998-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-08-01

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

KAERI results for BN600 full MOX benchmark (Phase 4)

International Nuclear Information System (INIS)

Lee, Kibog Lee

2003-01-01

The purpose of this document is to report the results of KAERI's calculation for the Phase-4 of BN-600 full MOX fueled core benchmark analyses according to the RCM report of IAEA CRP Action on U pdated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. T he BN-600 full MOX core model is based on the specification in the document, F ull MOX Model (Phase4. doc ) . This document addresses the calculational methods employed in the benchmark analyses and benchmark results carried out by KAERI

Design of the MOX fuel fabrication facility

International Nuclear Information System (INIS)

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

2001-01-01

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

Design of the MOX fuel fabrication facility

Energy Technology Data Exchange (ETDEWEB)

Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

2001-07-01

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

Photodissociation of C3H5Br and C4H7Br at 234 nm

International Nuclear Information System (INIS)

Kim, Hyun Kook; Paul, Dababrata; Hong, Ki Ryong; Cho, Ha Na; Kim, Tae Kyu; Lee, Kyoung Seok

2012-01-01

The photodissociation dynamics of cyclopropyl bromide (C-3H 5 Br) and cyclobutyl bromide (C 4 H 7 Br) at 234 nm was investigated. A two-dimensional photofragment ion-imaging technique coupled with a [2+1] resonance enhanced multiphoton ionization scheme was utilized to obtain speed and angular distributions of the nascent Br( 2 P 3/2 ) and Br*( 2 P 1/2 ) atoms. The recoil anisotropies for the Br and Br* channels were measured to be βBr = 0.92 ± 0.03 and βBr* = 1.52 ± 0.04 for C 3 H 5 Br and βBr = 1.10 ± 0.03 and βBr* = 1.49 ± 0.05 for C 4 H 7 Br. The relative quantum yield for Br was found to be ΦBr = 0.13 ± 0.03 and for C 3 H 5 Br and C 4 H 7 Br, respectively. The soft radical limit of the impulsive model adequately modeled the related energy partitioning. The nonadiabatic transition probability from the 3A' and 4A' potential energy surfaces was estimated and discussed

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

A risk-informed evaluation of MOX fuel loading in PWRS

International Nuclear Information System (INIS)

Lyman, E.S.

2001-01-01

The full text follows: The U.S. Department of Energy (DOE) has signed a contract with Duke Cogema Stone and Webster (DCS) for fabrication of mixed-oxide (MOX) fuel and irradiation of the MOX fuel at the Catawba and McGuire pressurized-water reactors (PWRs), operated by Duke Power. The first load of MOX fuel is scheduled for 2007. In order to use MOX in these plants, Duke Power will have to apply to the Nuclear Regulatory Commission (NRC) for amendments to their operating licenses. Until recently, there have been no numerical guidelines for determining the acceptability of license amendment requests. However, such guidelines are now at hand with the adoption in 1998 of NRC Regulatory Guide 1.174, which defines a maximum value for the permissible increase in risk to the public resulting from a proposed change to a nuclear plant's licensing basis (LB). The substitution of MOX fuel for low-enriched uranium (LEU) fuel in LWRs will have an impact on risk to the public that will require regulatory evaluation. One of the major differences is that use of MOX will increase the inventories of plutonium and minor actinides in the reactor core, thereby increasing the source term for certain severe accidents, such as a core melt with early containment failure or a spent fuel pool drain-down. The goal of this paper is to quantitatively evaluate the increase in risk associated with the greater actinide source term in MOX-fueled reactors, and to compare this increase with RG 1.174 guidelines. Standard computer programs (SCALE and MACCS2) are used to estimate the increase in severe accident risk to the public associated with the DCS plan to use 40% cores of weapons-grade MOX fuel. These values are then compared to the RG 1.174 acceptance criteria, using publicly available risk information. Since RG 1.174 guidelines are based on the assumption that severe accident source terms are not affected by LB changes, the RG 1.174 formalism must be modified for this case. A similar

Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

International Nuclear Information System (INIS)

Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

2000-03-01

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

Plasminogen activator inhibitor-1 4G/5G and the MTHFR 677C/T polymorphisms and susceptibility to polycystic ovary syndrome: a meta-analysis.

Science.gov (United States)

Lee, Young Ho; Song, Gwan Gyu

2014-04-01

The aim of this study was to explore whether the plasminogen activator inhibitor-1 (PAI-1) 4G/5G and the methylenetetrahydrofolate reductase (MTHFR) 677C/T polymorphisms are associated with susceptibility to polycystic ovary syndrome (PCOS). Meta-analyses were conducted to determine the association between the PAI-1 4G/5G and MTHFR 677C/T polymorphisms and PCOS using: (1) allele contrast (2) homozygote contrast, (3) recessive, and (4) dominant models. For meta-analysis, nine studies of the PAI-1 4G/5G polymorphism with 2384 subjects (PCOS, 1615; controls, 769) and eight studies of the MTHFR 677C/T polymorphism with 1270 study subjects were included. Meta-analysis of all study subjects showed no association between PCOS and the PAI-1 4G allele (OR=0.949, 95% CI=0.671-1.343, p=0.767). Stratification by ethnicity, however, indicated a significant association between the PAI-1 4G allele and PCOS in Turkish and Asian populations (OR=0.776, 95% CI=0.602-0.999, p=0.049; OR=1.749, 95% CI=1.297-2.359, p=2.5×10(-5) respectively). In addition, meta-analysis indicated an association between PCOS and the PAI-1 4G4G+4G5G genotype in Europeans (OR=1.406, 95% CI=1.025-1.928, p=0.035). However, meta-analysis of all study subjects showed no association between PCOS and the MTHFR 677T allele (OR=0.998, 95% CI=0.762-1.307, p=0.989), including Europeans (OR=0.806, 95% CI=0.610-1.063, p=0.126). Meta-analysis showed no association between PCOS and the MTHFR 677C/T polymorphism using homozygote contrast, and recessive and dominant models. In conclusion, meta-analysis suggests the PAI-1 4G/5G polymorphism is associated with susceptibility to PCOS in European, Turkish, and Asian populations, but the MTHFR 677C/T polymorphism is not associated with susceptibility to PCOS in Europeans. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

Plutonium - out of the stockpile and into the MOX market

International Nuclear Information System (INIS)

Edwards, J.; Hexter, B.C.; Powell, D.J.

1993-01-01

Reducing the risks associated with growing stocks of plutonium is just one of the factors behind the manufacture of mixed oxide (MOX) fuel. A United Kingdom collaboration, described here, has recently taken the first steps into the market place for MOX. (Author)

Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

International Nuclear Information System (INIS)

Takeda, Toshikazu

2001-01-01

Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

International Nuclear Information System (INIS)

Hirano, Fumio; Sato, Seichi; Kozaki, Tamotsu

2012-01-01

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO 2 spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO 2 spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80degC is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4-1.6, which is significantly lower than 4.0 for 45 GWd-UO 2 . (author)

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.

Development of a fresh MOX fuel transport package for disposition of weapons plutonium

International Nuclear Information System (INIS)

Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

1998-01-01

The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

Technical design considerations in the provision of a commercial MOX plant

International Nuclear Information System (INIS)

Elliott, M.F.

1997-01-01

The Sellafield MOX Plant (SMP) has a design production target of 120 t/year Heavy Metal of mixed uranium dioxide and plutonium dioxided (MOX) fuel. It will have the capability to produce fuel with fissile enrichments up to 10%. The feed materials are those arising from reprocessing operations on the Sellafield site, although the plant also has the capability to receive and process plutonium from overseas reprocessing plants. The ability to produce 10% enriched fuels, together with the requirement to use high burn-up feed has posed a number of design challenges to prevent excessive powder temperatures within the plant. As no stimulants are available to represent the heat generating nature of plutonium powders, it is difficult to prove equipment design by experiment. Extensive use has therefore been made of finite element analysis techniques. The requirement to process material of low burn-up (i.e. high fissile enrichment) has also impacted on equipment design in order to ensure that criticality limits are not exceeded. This has been achieved where possible by 'safe by geometry' design and, where appropriate, by high integrity protection systems. SMP has been designed with a high plant availability but at minimum cost. The requirement to minimize cost has meant that high availability must be obtained with the minimum of equipment. This had led to major challenges for equipment designers in terms of both the reliability and also the maintainability of equipment. Extensive use has been made of theoretical modelling techniques which have given confidence that plant throughput can be achieved. (author). 1 fig

Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

Status of irradiation testing and PIE of MOX (Pu-containing) fuel

International Nuclear Information System (INIS)

Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

1995-01-01

This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

Safety-related investigations on power distribution in MOX fuel elements in LWR cores

International Nuclear Information System (INIS)

Kramer, E.; Langenbuch, S.

1991-01-01

For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

NMR structural studies of the ionizing radiation adduct 7-hydro-8-oxodeoxyguanosine (8-oxo-7H-dG) opposite deoxyadenosine in a DNA duplex. 8-oxo-7H-dG(syn)·dA(anti) alignment at lesion site

International Nuclear Information System (INIS)

Kouchakdjian, M.; Patel, D.J.; Bodepudi, V.; Shibutani, S.; Eisenberg, M.; Johnson, F.; Grollman, A.P.

1991-01-01

Proton NMR studies are reported on the complementary d(C1-C2-A3-C4-T5-A6-oxo-G7-T8-C9-A10-C11-C12)·d(G13-G14-T15-G16-A17-A18-T19-A20-G21-T22-G23-G24) dodecanucleotide duplex (designated 8-oxo-7H-dG·dA 12-mer), which contains a centrally located 7-hydro-8-oxodeoxyguanosine (8-oxo-7H-dG) residue, a group commonly found in DNA that has been exposed to ionizing radiation or oxidizing free radicals. From the NMR spectra it can be deduced that this moiety exists as two tautomers, or gives rise to two DNA conformations, that are in equilibrium and that exchange slowly. The present study focuses on the major component of the equilibrium that originates in the 6,8-dioxo tautomer of 8-oxo-7H-dG. The authors have assigned the exchangeable NH1, NH7, and NH 2 -2 base protons located on the Watson-Crick and Hoogsteen edges of 8-oxo-7H-dG7 in the 8-oxo-7H-dG·dA 12-mer duplex, using an analysis of one- and two-dimensional nuclear Overhauser enhancement (NOE) data in H 2 O solution. They were able to detect a set of intra- and interstrand NOEs between protons (exchangeable and nonexchangeable) on adjacent residues in the d(A6-oxo-G7-T8)·d(A17-A18-T19) trinucleotide segment centered about the lesion site that establishes stacking of the oxo-dG7(syn)·dA(anti) pair between stable Watson-Crick dA6·dT19 and dT8·A17 base pairs with minimal perturbation of the helix. The structural studies demonstrate that 8-oxo-7H-dG(syn)·dA(anti) forms a stable pair in the interior of the helix, providing a basis for the observed incorporation of dA opposite 8-oxo-7H-dG when readthrough occurs past this oxidized nucleoside base

Atmospheric histories and growth trends of C4F10, C5F12, C6F14, C7F16 and C8F18

Directory of Open Access Journals (Sweden)

R. F. Weiss

2012-05-01

Full Text Available Atmospheric observations and trends are presented for the high molecular weight perfluorocarbons (PFCs: decafluorobutane (C4F10, dodecafluoropentane (C5F12, tetradecafluorohexane (C6F14, hexadecafluoroheptane (C7F16 and octadecafluorooctane (C8F18. Their atmospheric histories are based on measurements of 36 Northern Hemisphere and 46 Southern Hemisphere archived air samples collected between 1973 to 2011 using the Advanced Global Atmospheric Gases Experiment (AGAGE "Medusa" preconcentration gas chromatography-mass spectrometry systems. A new calibration scale was prepared for each PFC, with estimated accuracies of 6.8% for C4F10, 7.8% for C5F12, 4.0% for C6F14, 6.6% for C7F16 and 7.9% for C8F18. Based on our observations the 2011 globally averaged dry air mole fractions of these heavy PFCs are: 0.17 parts-per-trillion (ppt, i.e., parts per 1012 for C4F10, 0.12 ppt for C5F12, 0.27 ppt for C6F14, 0.12 ppt for C7F16 and 0.09 ppt for C8F18. These atmospheric mole fractions combine to contribute to a global average radiative forcing of 0.35 mW m−2, which is 6% of the total anthropogenic PFC radiative forcing (Montzka and Reimann, 2011; Oram et al., 2012. The growth rates of the heavy perfluorocarbons were largest in the late 1990s peaking at 6.2 parts per quadrillion (ppq, i.e., parts per 1015 per year (yr for C4F10, at 5.0 ppq yr−1 for C5F12 and 16.6 ppq yr−1 for C6F14 and in the early 1990s for C7F16 at 4.7 ppq yr−1 and in the mid 1990s for C8F18 at 4.8 ppq yr−1. The 2011 globally averaged mean atmospheric growth rates of these PFCs are subsequently lower at 2.2 ppq yr−1 for C4F10, 1.4 ppq yr−1 for C5F12, 5.0 ppq yr−1 for C6F14, 3.4 ppq yr−1 for C7F16 and 0.9 ppq yr−1 for C8F18. The more recent slowdown in the growth rates suggests that emissions are declining as compared to the 1980s and 1990s.

Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

International Nuclear Information System (INIS)

Omori, R.

2000-01-01

We have continued theoretical and experimental studies on laser manipulation of nuclear fuel particles, such as UO 2 , PuO 2 and ThO 2 , In this paper, we investigate the applicability of the collection of MOX particles floating in air using radiation pressure of a laser light; some preliminary results are shown. This technique will be useful for removal and confinement of MOX particles being transported by air current or dispersed in a cell box. First, we propose two types of principles for collecting MOX particles. Second, we show some experimental results, Third, we show numerical results of radiation pressure exerted on submicrometer-sized UO 2 particles using Generalized Lorentz-Mie theory. Because optical constants of UO 2 are similar to those of MOX fuel particles, it seems that calculation results obtained hold for MOX fuel particles. 2. Principles of collecting MOX fuel particles using radiation pressure (authors)

Instrumentation and procedures for moisture corrections to passive neutron coincidence counting assays of bulk PuO2 and MOX powders

International Nuclear Information System (INIS)

Stewart, J.E.; Menlove, H.O.; Ferran, R.R.; Aparo, M.; Zeppa, P.; Troiani, F.

1993-05-01

For passive neutron-coincidence-counting verification measurements of PuO 2 and MOX powder, assay biases have been observed that result from moisture entrained in the sample. This report describes a unique set of experiments in which MOX samples, with a range of moisture concentrations, were produced and used to calibrate and evaluate two prototype moisture monitors. A new procedure for moisture corrections to PuO 2 and MOX verification measurements yields MOX assays accurate to 1.5% (1σ) for 0.6- and 1.1-kg samples. Monte Carlo simulations were used to extend the measured moisture calibration data to higher sample masses. A conceptual design for a high-efficiency neutron coincidence counter with improved sensitivity to moisture is also presented

Bi-Modal Model for Neutron Emissions from PuO{sub 2} and MOX Holdup

Energy Technology Data Exchange (ETDEWEB)

Menlove, Howard; Lafleur, Adrienne [Los Alamos National Laboratory, Safeguard Science and Technology Group, NEN-1, MS E540, Los Alamos, NM, 87545 (United States)

2015-07-01

O{sub 2} particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm{sup 3}, but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O{sub 2} in the PuO{sub 2} as well as any impurity contamination such as H{sub 2}O. To obtain a more quantitative estimate of

Total Syntheses of (-)-Mersicarpine, (-)-Scholarisine G, (+)-Melodinine E, (-)-Leuconoxine, (-)-Leuconolam, (-)-Leuconodine A, (+)-Leuconodine F, and (-)-Leuconodine C: Self-Induced Diastereomeric Anisochronism (SIDA) Phenomenon for Scholarisine G and Leuconodines A and C.

Science.gov (United States)

Xu, Zhengren; Wang, Qian; Zhu, Jieping

2015-05-27

Enantioselective total syntheses of title natural products from a common cyclohexenone derivative (S)-18 were reported. Ozonolysis of (S)-18 afforded a stable diketo ester (R)-17 that was subsequently converted to two skeletally different natural products, i.e., (-)-mersicarpine (8) with a [6.5.6.7] fused tetracyclic ring system and (-)-scholarisine G (9) with a [6.5.6.6.5] fused pentacyclic skeleton, respectively. The postcyclization diversification was realized by taking advantage of the facile conversion of (+)-melodinine E (6) to N-acyliminium ion 7, from which a hydroxy group was selectively introduced to the C6, C7, C10 and the central C21 position of diazafenestrane system, leading to (-)-leuconodine A (11), (+)-leuconodine F (12), (-)-scholarisine G (9), (-)-leuconodine C (13), and skeletally different (-)-leuconolam (5). Furthermore, an unprecedented non-natural oxabridged oxadiazafenestrane 68 was formed by oxidation of (+)-melodinine E (6). During the course of this study, a strong self-induced diastereomeric anisochronism (SIDA) phenomenon was observed for scholarisine G (9), leuconodines A (11) and C (13). X-ray structures of both the racemic and the enantiopure natural products 9, 11, and 13 were obtained. The different crystal packing of these two forms nicely explained the chemical shift differences observed in the (1)H NMR spectra of the racemic and the enantio-enriched compounds in an achiral environment.

Adjacent-Vertex-Distinguishing Total Chromatic Numbers of C5 ∨ Kt%图C5∨Kt的邻点可区别全色数

Institute of Scientific and Technical Information of China (English)

张芳红; 王治文; 陈祥恩

2012-01-01

设f是图G的一个正常全染色.对任意x∈V(G),令C(x)表示与点x相关联的边的颜色以及点x的颜色所构成的集合.若对任意uv∈E(G),有C(u)≠C(v),则称f是图G的一个邻点可区别全染色.对一个图G进行邻点可区别全染色所需的最少的颜色的数目称为G的邻点可区别全色数,记为xat(G).用C5∨Kt表示长为5的圈与t阶完全图的联图.讨论了c5∨Kt的邻点可区别全色数.利用正多边形的对称性构造染色以及组合分析的方法,得到了当t是大于等于3的奇数以及t是偶数且2≤t≤22时,xat(C5∨Kt)＝t+6,当t是偶数且t≥24时,Xat(C5∨Kt)＝t+7.%Let / be a proper total coloring of G. For each x ∈V(G),let C(x) denote the set of all colors of the edhes incident with x and the color of x. If (V) uv ∈E(G),we have C(u) ≠ C(v),then / is called an adjacent vertex distinguishing total coloring of G. The minimum number k for which there exists an adjacent vertex distinguishing total coloring of G using k colors is called the adjacent vertex distinguishing total chromatic number of G and denoted by Xat(G). Let C5 V Kt be the join of the cycle C5 of order 5 and the complete graph Kt of order t. In this paper,we discuss adjacent-vertex-distinguishing total chromatic numbers of C5 V Kt. By using symmetry of regular polygons to construct coloring,and methods of combinatorial analysis,we obtained that for t is odd with t ≥ 3 or t is even with 2 ≤ t ≤ 22,we have Xat(C5,V Kt) = t + 6; for t is even with t ≥ 24,we have Xat(C5 V Kt) = t + 7.

A programmatic approach for implementing MOX fuel operation in advanced and existing boiling water reactors

International Nuclear Information System (INIS)

Ehrlich, E.H.; Knecht, P.D.; Shirley, N.C.; Wadekamper, D.C.

1996-01-01

This paper describes a programmatic overview of the elements and issues associated with MOX fuel utilization. Many of the dominant considerations and integrated relationships inherent in initiating MOX fuel utilization in BWRs or the ABWR with partial or full MOX core designs are discussed. The most significant considerations in carrying out a MOX implementation program, while achieving commercially desirable fuel cycles and commercially manageable MOX fuel fabrication, testing, qualification, and licensing support activities, are described. The impact of politics and public influences and the necessary role of industry and government contributions are also discussed. (J.P.N.)

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

Purification of the active C5a receptor from human polymorphonuclear leukocytes as a receptor - G sub i complex

Energy Technology Data Exchange (ETDEWEB)

Rollins, T.E.; Siciliano, S.; Kobayashi, S.; Cianciarulo, D.N.; Bonilla-Argudo, V.; Collier, K.; Springer, M.S. (Merck Sharp and Dohme Research Lab., Rahway, NJ (United States))

1991-02-01

The authors have isolated, in an active state, the C5a receptor from human polymorphonuclear leukocytes. The purification was achieved in a single step using a C5a affinity column in which the C5a molecule was coupled to the resin through its N terminus. The purified receptor, like the crude solubilized molecule, exhibited a single class of high-affinity binding sites with a K{sub d} of 30 pM. Further, the binding of C5a retained its sensitivity to guanine nucleotides, implying that the purified receptor contained a guanine nucleotide-binding protein (G protein). SDS/PAGE revealed the presence of three polypeptides with molecular masses of 42, 40, and 36 kDa, which were determined to be the C5a-binding subunit and the {alpha} and {beta} subunits of G{sub i}, respectively. The 36- and 40-kDa polypeptides were identified by immunoblotting and by the ability of pertussis toxin to ADP-ribosylate the 40-kDa molecule. These results confirm their earlier hypothesis that the receptor exists as a complex with a G protein in the presence or absence of C5a. The tight coupling between the receptor and G protein should make possible the identification of the G protein(s) involved in the transduction pathways used by C5a to produce its many biological effects.

ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-08-01

A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

Structure of human ubiquitin-conjugating enzyme E2 G2 (UBE2G2/UBC7)

International Nuclear Information System (INIS)

Arai, Ryoichi; Yoshikawa, Seiko; Murayama, Kazutaka; Imai, Yuzuru; Takahashi, Ryosuke; Shirouzu, Mikako; Yokoyama, Shigeyuki

2006-01-01

The crystal structure of human UBE2G2/UBC7 was solved at 2.56 Å resolution. The superimposition of UBE2G2 on UbcH7 in a c-Cbl–UbcH7–ZAP70 ternary complex suggested that the two loop regions of UBE2G2 interact with the RING domain in a similar way as UbcH7. The human ubiquitin-conjugating enzyme E2 G2 (UBE2G2/UBC7) is involved in protein degradation, including a process known as endoplasmic reticulum-associated degradation (ERAD). The crystal structure of human UBE2G2/UBC7 was solved at 2.56 Å resolution. The UBE2G2 structure comprises a single domain consisting of an antiparallel β-sheet with four strands, five α-helices and two 3 10 -helices. Structural comparison of human UBE2G2 with yeast Ubc7 indicated that the overall structures are similar except for the long loop region and the C-terminal helix. Superimposition of UBE2G2 on UbcH7 in a c-Cbl–UbcH7–ZAP70 ternary complex suggested that the two loop regions of UBE2G2 interact with the RING domain in a similar way to UbcH7. In addition, the extra loop region of UBE2G2 may interact with the RING domain or its neighbouring region and may be involved in the binding specificity and stability

The development of B.N.F.L.'S MOX fuel supply business

International Nuclear Information System (INIS)

Edwards, J.; Brown, C.; Marshall, S.J.; Connell, M.; Thompson, H.

1998-01-01

In 1990 BNFL developed a strategy to become one of the world leading MOX fuel suppliers. This strategy involved the design, construction and operation of a small scale demonstration plant known as the MOX Demonstration Facility (MDF) and a large scale facility known as the Sellafield MOX Plant (SMP). To support the development of these facilities, BNFL developed a new MOX fuel fabrication process known as the Short Binderless Route (SBR). Since the 1990 decision was made, the company has successfully built, commissioned and operated the MDF, and has designed, built and is in the process of commissioning the 120 t(HM)/year SMP. The scale of the business has thus developed significantly and the direction and prospects for the future of the business are clear and well understood, with the focus being on the use of BNFL technology to produce quality MOX fuel to meet customers' requirements. This paper reviews the development of BNFL's MOX business and describes the technology being used in the state of the art SMP. The paper also explains the approach taken to commission the plant and how key safety features have been incorporated into the design. Up to date information on the performance of Short Binderless Route fuel is provided, and the future development of the business is discussed. (author)

The human polynucleotide kinase/phosphatase (hPNKP) inhibitor A12B4C3 radiosensitizes human myeloid leukemia cells to Auger electron-emitting anti-CD123 111In-NLS-7G3 radioimmunoconjugates

International Nuclear Information System (INIS)

Zereshkian, Arman; Leyton, Jeffrey V.; Cai, Zhongli; Bergstrom, Dane; Weinfeld, Michael; Reilly, Raymond M.

2014-01-01

Introduction: Leukemia stem cells (LSCs) are believed to be responsible for initiating and propagating acute myeloid leukemia (AML) and for causing relapse after treatment. Radioimmunotherapy (RIT) targeting these cells may improve the treatment of AML, but is limited by the low density of target epitopes. Our objective was to study a human polynucleotide kinase/phosphatase (hPNKP) inhibitor that interferes with DNA repair as a radiosensitizer for the Auger electron RIT agent, 111 In-NLS-7G3, which recognizes the CD123 + /CD131 - phenotype uniquely displayed by LSCs. Methods: The surviving fraction (SF) of CD123 + /CD131 - AML-5 cells exposed to 111 In-NLS-7G3 (33–266 nmols/L; 0.74 MBq/μg) or to γ-radiation (0.25-5 Gy) was determined by clonogenic assays. The effect of A12B4C3 (25 μmols/L) combined with 111 In-NLS-7G3 (16–66 nmols/L) or with γ-radiation (0.25–2 Gy) on the SF of AML-5 cells was assessed. The density of DNA double-strand breaks (DSBs) in the nucleus was measured using the γ-H2AX assay. Cellular dosimetry was estimated based on the subcellular distribution of 111 In-NLS-7G3 measured by cell fractionation. Results: Binding of 111 In-NLS-7G3 to AML-5 cells was reduced by 2.2-fold in the presence of an excess (1 μM) of unlabeled NLS-7G3, demonstrating specific binding to the CD123 + /CD131 - epitope. 111 In-NLS-7G3 reduced the SF of AML-5 cells from 86.1 ± 11.0% at 33 nmols/L to 10.5 ± 3.6% at 266 nmols/L. Unlabeled NLS-7G3 had no significant effect on the SF. Treatment of AML-5 cells with γ-radiation reduced the SF from 98.9 ± 14.9% at 0.25 Gy to 0.03 ± 0.1% at 5 Gy. A12B4C3 combined with 111 In-NLS-7G3 (16–66 nmols/L) enhanced the cytotoxicity up to 1.7-fold compared to treatment with radioimmunoconjugates alone and was associated with a 1.6-fold increase in DNA DSBs in the nucleus. A12B4C3 enhanced the cytotoxicity of γ-radiation (0.25–0.5 Gy) on AML-5 cells by up to 1.5-fold, and DNA DSBs were increased by 1.7-fold. Exposure to

Development of a reference scheme for MOX lattice physics calculations

International Nuclear Information System (INIS)

Finck, P.J.; Stenberg, C.G.; Roy, R.

1998-01-01

The US program to dispose of weapons-grade Pu could involve the irradiation of mixed-oxide (MOX) fuel assemblies in commercial light water reactors. This will require licensing acceptance because of the modifications to the core safety characteristics. In particular, core neutronics will be significantly modified, thus making it necessary to validate the standard suites of neutronics codes for that particular application. Validation criteria are still unclear, but it seems reasonable to expect that the same level of accuracy will be expected for MOX as that which has been achieved for UO 2 . Commercial lattice physics codes are invariably claimed to be accurate for MOX analysis but often lack independent confirmation of their performance on a representative experimental database. Argonne National Laboratory (ANL) has started implementing a public domain suite of codes to provide for a capability to perform independent assessments of MOX core analyses. The DRAGON lattice code was chosen, and fine group ENDF/B-VI.04 and JEF-2.2 libraries have been developed. The objective of this work is to validate the DRAGON algorithms with respect to continuous-energy Monte Carlo for a suite of realistic UO 2 -MOX benchmark cases, with the aim of establishing a reference DRAGON scheme with a demonstrated high level of accuracy and no computing resource constraints. Using this scheme as a reference, future work will be devoted to obtaining simpler and less costly schemes that preserve accuracy as much as possible

Chemical Synthesis of a 5'-Terminal TMG-Capped Triribonucleotide m(3)(2,2,7)G(5)(')pppAmpUmpA of U1 RNA.

Science.gov (United States)

Sekine, Mitsuo; Kadokura, Michinori; Satoh, Takahiko; Seio, Kohji; Wada, Takeshi; Fischer, Utz; Sumpter, Vicki; Lührmann, Reinhard

1996-06-26

The 5'-terminal TMG-capped triribonucleotide, m(3)(2,2,7)G(5)(')pppAmpUmpA, has been synthesized by condensation of an appropriately protected triribonucleotide derivative of ppAmpUmpA with a new TMG-capping reagent. During this total synthesis, it was found that the regioselective 2'-O-methylation of 3',5'-O-(1,1,3,3-tetraisopropyldisiloxane-1,3-diyl)-N-(4-monomethoxytrityl)adenosine was achieved by use of MeI/Ag(2)O without affecting the base moiety. A new route to 2-N,2-N-dimethylguanosine from guanosine via a three-step reaction has also been developed by reductive methylation using paraformaldehyde and sodium cyanoborohydride. These key intermediates were used as starting materials for the construction of a fully protected derivative of pAmpUmpA and a TMG-capping reagent of Im-pm(3)(2,2,7)G. The target TMG-capped tetramer, m(3)(2,2,7)G(5)(')pppAmpUmpA, was synthesized by condensation of a partially protected triribonucleotide 5'-terminal diphosphate species, ppA(MMTr)mpUmpA, with Im-pm(3)(2,2,7)G followed by treatment with 80% acetic acid. The structure of m(3)(2,2,7)G(5)(')pppAmpUmpA was characterized by (1)H and (31)P NMR spectroscopy as well as enzymatic assay using snake venom phosphodiesterase, calf intestinal phosphatase, and nuclease P1.

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

The association of factor V G1961A (factor V Leiden), prothrombin G20210A, MTHFR C677T and PAI-1 4G/5G polymorphisms with recurrent pregnancy loss in Bosnian women.

Science.gov (United States)

Jusić, Amela; Balić, Devleta; Avdić, Aldijana; Pođanin, Maja; Balić, Adem

2018-08-01

Aim To investigate association of factor V Leiden, prothrombin G20210A, MTHFR C677T and PAI-1 4G/5G polymorphisms with recurrent pregnancy loss in Bosnian women. Methods A total of 60 women with two or more consecutive miscarriages before 20 weeks of gestation with the same partners and without history of known causes or recurrent pregnancy loss were included. A control group included 80 healthy women who had one or more successful pregnancies without history of any complication which could be associated with miscarriages. Genotyping of factor V Leiden, prothrombin G20210A, MTHFR C677T and PAI-1 4G/5G polymorphisms were performed by polymerase chain reaction/restriction fragments length polymorphism method (PCR/RFLP). Results Both factor V Leiden and MTHFR C677T polymorphisms were significantly associated with recurrent pregnancy loss (RPL) in Bosnian women while prothrombin G20210A and PAI-1 4G/5G polymorphisms did not show strongly significant association. Conclusion The presence of thrombophilic polymorphisms may predispose women to recurrent pregnancy loss. Future investigation should be addressed in order to find when carriers of those mutations, polymorphisms should be treated with anticoagulant therapy. Copyright© by the Medical Assotiation of Zenica-Doboj Canton.

Mol 7C/6; Mol 7C/6

Energy Technology Data Exchange (ETDEWEB)

Aberle, J.; Schleisiek, K.; Schmuck, I.; Schmidt, L.; Romer, O.; Weih, G.

1995-08-01

The Mol 7C/6 coolant blockage experiment in the Belgian BR2 reactor yielded results different from Mol 7C experiments with low burnup pins: At 10% burnup local failure is not self-limiting, but requires active systems for detection and scram. The Mol 7C series was finished in 1991. In each of the test bundles Mol 7C/4, /5 and /6, 30 Mk I pins pre-irradiated in KNK II were used. The central blockage consisted of enriched UO{sub 2} covering 30 percent of the bundle cross-section, with a height of 40 mm. The most important system for timely detection of coolant blockages of the type studied in Mol 7C/6 is based on DND. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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

International Nuclear Information System (INIS)

Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O'Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

1998-03-01

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

Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

International Nuclear Information System (INIS)

Ludwig, S.B.; Michelhaugh, R.D.; Pope, R.B.

1997-12-01

This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies

Validation of the TUBRNP model with the radial distribution of plutonium in MOX fuel measured by SIMS and EPMA

Energy Technology Data Exchange (ETDEWEB)

O` Carroll, C; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

1997-08-01

The new model TUBRNP (TRANSURANUS burnup) predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of plutonium. Comparisons between measurements and the prediction of the TUBRNP model have been made for UO{sub 2} LWR fuels: they were found to be in excellent agreement and it is seen that TUBRNP is a marked improved on previous models. A powerful techniques for the characterization of irradiation fuel is Electron Probe Microanalysis (EPMA). Uranium, plutonium and fission product distributions can be analysed quantitatively. A complement, providing isotopic information with a lateral resolution comparable to EPMA, is secondary ion mass spectrometry (SIMS). Recently, the technique has been successfully applied for the measurement of the radial distribution of plutonium isotopes in irradiated nuclear fuel pins. The extension of the TUBRNP model to mixed oxide fuels seems to be the natural step to take. In MOX fuels the picture is greatly complicated by the presence of the (U, Pu)O{sub 2} agglomerates. The rim effect referred to above may be masked by the high concentrations of plutonium in the bulk of the fuel. A detailed investigation of a number of MOX fuel samples has been made using the TUBRNP model. Results are presented for a range of fuels with different enrichment and burnup. Through its participation in the PRIMO and DOMO programmes, PSI in conjunction with the Institute for Transuranium Elements had the opportunity to validate the new theoretical model TUBRNP. The authors with therefore to express their thanks to the organizers and to the numerous European and Japanese organizations which have supported these two international programmes on MOX fuel behavior. 7 refs, 9 figs, 3 tabs.

Analysis of a MOX-UO2 interface by the method of characteristics

International Nuclear Information System (INIS)

Chetaine, A.; Erradi, L.; Sanchez, R.; Zmijarevic, I.; Aniel-Buchheit, S.

2005-01-01

In the last few years many studies have been done to improve the ability of core reactors (PWR and BWR) to burn Plutonium fuel, either in mixed UO 2 /MOX pattern or full MOX pattern. The analysis of a MOX-UO 2 interface with the method of characteristics has been carried out. Comparisons with Monte Carlo and collision-probability calculations show that our results are in good agreement with those obtained by reference methods and qualify the method of characteristic as a reliable technique for such calculations. (authors)

Characterization of acid flux in osteoclasts from patients harboring a G215R mutation in ClC-7

DEFF Research Database (Denmark)

Henriksen, Kim; Gram, Jeppe; Neutzsky-Wulff, Anita Vibsig

2008-01-01

-I, calcium release, and pit area when comparing to age and sex matched controls. In addition, the ADOII osteoclasts showed no differences in actin ring formation. Finally, V-ATPase and chloride channel inhibitors completely abrogated the H(+) and Cl(-) driven acidification. Finally, the acid influx...... mutation in ClC-7 (G215R) were isolated, and used these to investigate bone resorption measured by CTX-I, calcium release and pit scoring. The actin cytoskeleton of the osteoclasts was also investigated. ClC-7 enriched membranes from the osteoclasts were isolated, and used to test acidification rates...... in the presence of a V-ATPase and a chloride channel inhibitor, using a H(+) and Cl(-) driven approach. Finally, acidification rates in ClC-7 enriched membranes from ADOII osteoclasts and their corresponding controls were compared. Resorption by the G215R osteoclasts was reduced by 60% when measured by both CTX...

Altered Pre-mRNA Splicing Caused by a Novel Intronic Mutation c.1443+5G>A in the Dihydropyrimidinase (DPYS) Gene.

Science.gov (United States)

Nakajima, Yoko; Meijer, Judith; Zhang, Chunhua; Wang, Xu; Kondo, Tomomi; Ito, Tetsuya; Dobritzsch, Doreen; Van Kuilenburg, André B P

2016-01-12

Dihydropyrimidinase (DHP) deficiency is an autosomal recessive disease caused by mutations in the DPYS gene. Patients present with highly elevated levels of dihydrouracil and dihydrothymine in their urine, blood and cerebrospinal fluid. The analysis of the effect of mutations in DPYS on pre-mRNA splicing is hampered by the fact that DHP is primarily expressed in liver and kidney cells. The minigene approach can detect mRNA splicing aberrations using cells that do not express the endogenous mRNA. We have used a minigene-based approach to analyze the effects of a presumptive pre-mRNA splicing mutation in two newly identified Chinese pediatric patients with DHP deficiency. Mutation analysis of DPYS showed that both patients were compound heterozygous for a novel intronic mutation c.1443+5G>A in intron 8 and a previously described missense mutation c.1001A>G (p.Q334R) in exon 6. Wild-type and the mutated minigene constructs, containing exons 7, 8 and 9 of DPYS, yielded different splicing products after expression in HEK293 cells. The c.1443+5G>A mutation resulted in altered pre-mRNA splicing of the DPYS minigene construct with full skipping of exon 8. Analysis of the DHP crystal structure showed that the deletion of exon 8 severely affects folding, stability and homooligomerization of the enzyme as well as disruption of the catalytic site. Thus, the analysis suggests that the c.1443+5G>A mutation results in aberrant splicing of the pre-mRNA encoding DHP, underlying the DHP deficiency in two unrelated Chinese patients.

Experimental and Theoretical Studies of the Factors Affecting the Cycloplatination of the Chiral Ferrocenylaldimine (SC-[(η5-C5H5Fe{(η5-C5H4–C(H=N–CH(Me(C6H5}

Directory of Open Access Journals (Sweden)

Concepción López

2014-11-01

Full Text Available The study of the reactivity of the enantiopure ferrocenyl Schiff base (SC-[FcCH=N–CH(Me(C6H5] (1 (Fc = (η5-C5H5Fe(η5-C5H4 with cis-[PtCl2(dmso2] under different experimental conditions is reported. Four different types of chiral Pt(II have been isolated and characterized. One of them is the enantiomerically pure trans-(SC-[Pt{κ1-N[FcCH=N–CH(Me(C6H5]}Cl2(dmso] (2a in which the imine acts as a neutral N-donor ligand; while the other three are the cycloplatinated complexes: [Pt{κ2-C,N [(C6H4–N=CHFc]}Cl(dmso] (7a and the two diastereomers {(Sp,SC and (Rp,SC} of [Pt{κ2-C,N[(η5-C5H3–CH=N–{CH(Me(C6H5}]Fe(η5-C5H5}Cl(dmso] (8a and 9a, respectively. Isomers 7a-9a, differ in the nature of the metallated carbon atom [CPh (in 7a or CFc (in 8a and 9a] or the planar chirality of the 1,2-disubstituted ferrocenyl unit (8a and 9a. Reactions of 7a–9a with PPh3 gave [Pt{κ2-C,N[(C6H4–N=CHFc]}Cl(PPh3] (in 7b and the diastereomers (Sp,SC and (Rp,SC of [Pt{κ2-C,N[(η5-C5H3–CH=N–{CH(Me(C6H5}] Fe(η5-C5H5}Cl(PPh3] (8b and 9b, respectively. Comparative studies of the electrochemical properties and cytotoxic activities on MCF7 and MDA-MB231 breast cancer cell lines of 2a and cycloplatinated complexes 7b-9b are also reported. Theoretical studies based on DFT calculations have also been carried out in order to rationalize the results obtained from the cycloplatination of 1, the stability of the Pt(II complexes and their electrochemical properties.

Analysis of the Relationship between CGB5 155G/C Polymorphism and in vitro Fertilization-embryo Transfer Outcome (IVF-ET in the Iranian Population

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Bahareh Babaei Houlari

2018-01-01

Full Text Available Abstract Background: Successful pregnancy depends on the ability of the embryo to achieve appropriate extent of trophoblastic proliferation and invasion into maternal endometrium as well as, once implanted, to induce its own blood supply. Beta Human chorionic gonadotropin (β-hCG, enhances blastocyst implantation, uterine vascularization, and angiogenesis, as well as regulates maintenance of uterine quiescence and immunological adaptation during pregnancy. The β-subunit of hCG is encoded by CGB3, CGB6, CGB5, CGB7 and CGB8 genes. The aim of this study was to evaluate the association of CGB5-G/C polymorphism and the clinical outcomes in women who underwent IVF-ET procedures. Materials and Methods: A total of 200 patients undergoing IVF-ET (100 patients with positive and 100 patients with negative IVF-ET outcome were included in this study. Genotyping of CGB5 at -155G/C polymorphic site was performed by polymerase chain reaction-restriction fragment length polymorphism (PCR-RFLP. Statistical analysis was performed using the MedCalc software. Results: Our findings show that the CC genotype of the CGB5 -155G/C polymorphism is associated with decreased risk of IVF-ET failure (OR=0.29; 95%CI=0.1-0.85; p=0.02. However, the allelic distribution of the CGB5 -155G/C is not significantly different between two groups (χ2=1.46; p=0.22. Conclusion: The results of this study suggested that CGB5 (-155G/C CC genotype has a protective effect on IVF-ET outcome. More studies with larger sample sizes on different populations are necessary to elucidate the underlying mechanisms which can explain the associations found between the GGB5 gene polymorphisms and IVF-ET outcome.

Identification of nucleotides in the 5'UTR and amino acids substitutions that are essential for the infectivity of 5'UTR-NS5A recombinant of hepatitis C virus genotype 1b (strain Con1).

Science.gov (United States)

Li, Jinqian; Feng, Shengjun; Liu, Xi; Guo, Mingzhe; Chen, Mingxiao; Chen, Yiyi; Rong, Liang; Xia, Jinyu; Zhou, Yuanping; Zhong, Jin; Li, Yi-Ping

2018-05-01

Genotype 1b strain Con1 represents an important reference in the study of hepatitis C virus (HCV). Here, we aimed to develop an advanced infectious Con1 recombinant. We found that previously identified mutations A1226G/F1464L/A1672S/Q1773H permitted culture adaption of Con1 Core-NS5A (C-5A) recombinant containing 5'UTR and NS5B-3'UTR from JFH1 (genotype 2a), thus acquired additional mutations L725H/F886L/D2415G. C-5A containing all seven mutations (C-5A_7m) replicated efficiently in Huh7.5 and Huh7.5.1 cells and had an increased infectivity in SEC14L2-expressing Huh7.5.1 cells. Incorporation of Con1 NS5B was deleterious to C-5A_7m, however Con1 5'UTR was permissive but attenuated the virus. Nucleotides G1, A4, and G35 primarily accounted for the viral attenuation without affecting RNA translation. C-5A_7m was inhibited dose-dependently by simeprevir and daclatasvir, and substitutions at A4, A29, A34, and G35 conferred resistance to miR-122 antagonism. The novel Con1 5'UTR-NS5A recombinant, adaptive mutations, and critical nucleotides described here will facilitate future studies of HCV culture systems and virus-host interaction. Copyright © 2018 Elsevier Inc. All rights reserved.

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

Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

Energy Technology Data Exchange (ETDEWEB)

Odorowski, Mélina [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Jégou, Christophe, E-mail: christophe.jegou@cea.fr [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); De Windt, Laurent [MINES ParisTech, PSL Research University, Centre de Géosciences, 35 rue St Honoré, 77305 Fontainebleau (France); Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly [CEA/DEN/DTCD/SECM/LMPA, BP 17171, 30207 Bagnols-sur-Cèze Cedex (France); Martin, Christelle [Agence nationale pour la gestion des déchets radioactifs (Andra), DRD/CM, 1-7 rue Jean-Monnet, 92298 Châtenay-Malabry Cedex (France)

2016-01-15

Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 10{sup 9} Bq.g{sub MOX}{sup −1} reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·t{sub HM}{sup −1} after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O{sub 2}] < 1 ppm) for one year in carbonated water (10{sup −2} mol L{sup −1}). For each experiment, solution samples were taken over time and Eh and pH were monitored. The uranium in solution was assayed using a kinetic phosphorescence analyzer (KPA), plutonium and americium were analyzed by a radiochemical route, and H{sub 2}O{sub 2} generated by the water radiolysis was quantified by chemiluminescence. Surface characterizations were performed before and after leaching using Scanning Electron Microscopy (SEM), Electron Probe Microanalyzer (EPMA) and Raman spectroscopy. Solubility diagrams were calculated to support data discussion. The uranium releases from MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO{sub 2} reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO{sub 2} matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO{sub 2} grains were much more

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

International Nuclear Information System (INIS)

Kosaka, Yuji; Kurematsu, Shigeru; Kitagawa, Takaaki; Suzuki, Akihiro; Terai, Takayuki

2012-01-01

The temperature measurements of mixed oxide (MOX) and UO 2 fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO 2 fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel. (author)

Current Status of J-MOX Safeguards Design and Future Prospects

International Nuclear Information System (INIS)

Sampei, T.; Hiruta, K.; Shimizu, J.; Ikegame, K.

2015-01-01

The construction of JNFL MOX Fuel Fabrication Plant (J-MOX) is proceeding toward active test using uranium and MOX in July 2017, and completion of construction in October 2017. Although the construction schedule is largely impacted by progress of licencing, according to domestic law, JNFL is making every effort to get necessary permission of business licence and authorization of design and construction method as soon as possible. On the other hand, it is desirable that integrated safeguards approach is effective, efficient and consistent with J-MOX facility features. Discussion about the approach is going on among IAEA, Japan Safeguards Office (JSGO) and JNFL, and IAEA is planning to introduce the measures into the approach such as application of Near Real-Time Accountancy with frequent declaration from operator, Containment/Surveillance measures to storages, internal flow verification with 100%, random interim inspection (RII) and so on. RII scheme is intended to increase efficiency without compromising effectiveness and makes interruption of facility operation minimum. Also newly developed and improved safeguards equipment will be employed and it is possible to realize to increase credibility and efficiency of inspection by introduction of unattended/automatic safeguards equipment. Especially IAEA and JSGO share the development of non-destructive assay systems which meet the requirements from both parties. These systems will be jointly utilized at the flow verification, RII and PIV. JNFL will continue to provide enough design information in a timely manner toward early establishment of safeguards approach for J-MOX. Also JNFL will implement the coordination of installation and commissioning of safeguards equipment, and Design Information Verification activities for completion of construction in October 2017

Approach to customer qualification of the BNFL Sellafield Mox Plant

International Nuclear Information System (INIS)

Sullivan, P.

2003-01-01

BNFL started plutonium commissioning of its Sellafield MOX Plant (SMP) in December 2001, with the first MOX pellets being produced in May 2002. SMP was designed to manufacture a range of both PWR and BWR fuel types for a number of different customers. During commissioning and early MOX fuel manufacturing BNFL has been demonstrating its ability to both automatically manufacture and inspect MOX fuel to meet the requirements of different customers' specifications and fuel types. The qualification project consisted of common and project specific qualification. Common qualification was carried out to demonstrate BNFL could meet several customers' requirements during the same qualification test. Project specific qualification was carried out for one customer only as the fabrication or inspection equipment was specific to their fuel type. An example is the fuel assembly process. The reasons for BNFL carrying out common qualification were: - Develop a common qualified process to meet different customer specifications. - Minimise future qualifications prior to starting future fuel campaigns. - Ensure BNFL understands and effectively manages different customer requirements in SMP. BNFL has approached qualification of SMP systematically. Firstly the inspection system was qualified, and once completed the inspection system was then used in the qualification of the manufacturing process. (orig.)

LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

International Nuclear Information System (INIS)

Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

1986-09-01

The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

Safety and licensing of MOX versus UO2 for BWRs and PWRs: Aspects applicable for civilian and weapons grade Pu

International Nuclear Information System (INIS)

Goldstein, L.; Malone, J.

2000-01-01

This paper reviews the safety and licensing differences between MOX and UO 2 BWR and PWR cores. MOX produced from the normal recycle route and from weapons grade material are considered. Reload quantities of recycle MOX assemblies have been licensed and continue to operate safely in European LWRs. In general, the European MOX assemblies in a reload are 2 . These studies indicated that no important technical or safety related issues have evolved from these studies. The general specifications used by fuel vendors for recycled MOX fuel and core designs are as follows: MOX assemblies should be designed to minimize or eliminate local power peaking mismatches with co-resident and adjacently loaded UO 2 assemblies. Power peaking at the interfaces arises from different neutronic behavior between UO 2 and MOX assemblies. A MOX core (MOX and UO 2 or all-MOX assemblies) should provide cycle energy equivalent to that of an all-UO 2 core. This applies, in particular, to recycle MOX applications. An important consideration when burning weapons grade material is rapid disposition which may not necessarily allow for cycle energy equivalence. The reactivity coefficients, kinetics data, power peaking, and the worth of shutdown systems with MOX fuel and cores must be such to meet the design criteria and fulfill requirements for safe reactor operation. Both recycle and weapons grade plutonium are considered, and positive and negative impacts are given. The paper contrasts MOX versus UO 2 with respect to safety evaluations. The consequences of some transients/accidents are compared for both types of MOX and UO 2 fuel. (author)

Dynamic Monte Carlo transient analysis for the Organization for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) C5G7-TD benchmark

Energy Technology Data Exchange (ETDEWEB)

Shaukat, Nadeem; Ryu, Min; Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

2017-08-15

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

Altered Pre-mRNA Splicing Caused by a Novel Intronic Mutation c.1443+5G>A in the Dihydropyrimidinase (DPYS Gene

Directory of Open Access Journals (Sweden)

Yoko Nakajima

2016-01-01

Full Text Available Dihydropyrimidinase (DHP deficiency is an autosomal recessive disease caused by mutations in the DPYS gene. Patients present with highly elevated levels of dihydrouracil and dihydrothymine in their urine, blood and cerebrospinal fluid. The analysis of the effect of mutations in DPYS on pre-mRNA splicing is hampered by the fact that DHP is primarily expressed in liver and kidney cells. The minigene approach can detect mRNA splicing aberrations using cells that do not express the endogenous mRNA. We have used a minigene-based approach to analyze the effects of a presumptive pre-mRNA splicing mutation in two newly identified Chinese pediatric patients with DHP deficiency. Mutation analysis of DPYS showed that both patients were compound heterozygous for a novel intronic mutation c.1443+5G>A in intron 8 and a previously described missense mutation c.1001A>G (p.Q334R in exon 6. Wild-type and the mutated minigene constructs, containing exons 7, 8 and 9 of DPYS, yielded different splicing products after expression in HEK293 cells. The c.1443+5G>A mutation resulted in altered pre-mRNA splicing of the DPYS minigene construct with full skipping of exon 8. Analysis of the DHP crystal structure showed that the deletion of exon 8 severely affects folding, stability and homooligomerization of the enzyme as well as disruption of the catalytic site. Thus, the analysis suggests that the c.1443+5G>A mutation results in aberrant splicing of the pre-mRNA encoding DHP, underlying the DHP deficiency in two unrelated Chinese patients.

PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

Energy Technology Data Exchange (ETDEWEB)

Kyser, E.; King, W.; O' Rourke, P.

2012-07-26

Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

Performance evaluation of WDXRF as a process control technique for MOX fuel fabrication

International Nuclear Information System (INIS)

Pandey, A.; Khan, F.A.; Das, D.K.; Behere, P.G.; Afzal, Mohd

2015-01-01

This paper presents studies on Wavelength Dispersive X-Ray Fluorescence (WDXRF), as a powerful non destructive technique (NDT) for the compositional analysis of various types of MOX fuels. The sample has come after mixing and milling of UO 2 and PuO 2 powder for the estimation of plutonium, as a process control step of fabrication of (U, Pu)O 2 mixed oxide (MOX) fuel. For the characterization for heavy metal in various MOX fuel, a WDXRF method was established as a process control technique. The attractiveness of our system is that it can analyze the samples in solid form as well as in liquid form. The system is adapted in a glove box for handling of plutonium based fuels. The glove box adapted system was optimized with Uranium and Thorium based MOX sample before introduction of Pu. Uranium oxide and thorium oxide have been estimated in uranium thorium MOX samples. Standard deviation for the analysis of U 3 O 8 and ThO 2 were found to be 0.14 and 0.15 respectively. The results are validated against the conventional wet chemical methods of analysis. (author)

PAI-1 4G/5G and MTHFR C677T polymorphisms increased the accuracy of two prediction scores for the risk of acute lower extremity deep vein thrombosis.

Science.gov (United States)

Pop, Tudor Radu; Vesa, Ştefan Cristian; Trifa, Adrian Pavel; Crişan, Sorin; Buzoianu, Anca Dana

2014-01-01

This study investigates the accuracy of two scores in predicting the risk of acute lower extremity deep vein thrombosis. The study included 170 patients [85 (50%) women and 85 (50%) men] who were diagnosed with acute lower extremity deep vein thrombosis (DVT) with duplex ultrasonography. Median age was 62 (52.75; 72) years. The control group consisted of 166 subjects [96 (57.8%) women and 70 (42.2%) men], without DVT, matched for age (± one year) to those in the group with DVT. The patients and controls were selected from those admitted to the internal medicine, cardiology and geriatrics wards within the Municipal Hospital of Cluj-Napoca, Romania, between October 2009 and June 2011. Clinical, demographic and lab data were recorded for each patient. For each patient we calculated the prior risk of DVT using two prediction scores: Caprini and Padua. According to the Padua score only 93 (54.7%) patients with DVT had been at high risk of developing DVT, while 48 (28.9%) of controls were at high risk of developing DVT. When Padua score included PAI-1 4G/5G and MTHFR C677T polymorphisms, the sensitivity increased at 71.7%. Using the Caprini score, we determined that 147 (86.4%) patients with DVT had been at high risk of developing DVT, while 103 (62%) controls were at high risk of developing DVT. A Caprini score higher than 5 was the strongest predictor of acute lower extremity DVT risk. The Caprini prediction score was more sensitive than the Padua score in assessing the high risk of DVT in medical patients. PAI-1 4G/5G and MTHFR C677T polymorphisms increased the sensitivity of Padua score.

MOX fuel use as a back-end option: Trends, main issues and impacts on fuel cycle management

International Nuclear Information System (INIS)

Fukuda, K.; Choi, J.-S.; Shani, R.; Durpel, L. van den; Bertel, E.; Sartori, E.

2000-01-01

In the past decades while the FBIULWR fuel cycle concept was zealously being developed, MOX-fuel use in thermal reactors was taken as an alternative back-end policy option. However, the plutonium recycling with LWRs has evolved to industrial level, gaining high maturity through the incubative period while FBR deployment was envisaged. Today, MOX-fuel use in LWRs makes integral part of the fuel cycle for those countries relying on the recycling policy. Developments to improve the fuel cycle performance, including the minimisation of remaining wastes, and the reactor engineering aspects owing to MOX-fuel use, are continued. This paper jointly presented by IAEA and OECD/NEA brings an integrated overview on MOX use as a back-end policy, covering MOX fuel utilisation, fuel performance and technology, economics, licensing, MOX fuel trends in the coming decades. (author)

Fuel production for LWRs - MOX fuel aspects

International Nuclear Information System (INIS)

Deramaix, P.

2005-01-01

Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O 2 fuel pellets (MOX), fabricated to a large extent according to UO 2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO 2 . The industrial processes presently in use or planned are all based on a mechanical blending of UO 2 and PuO 2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO 2 , PuO 2 ) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO 2 containing about 30% PuO 2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO 2 to obtain the specified Pu content. After mixing, the (U, Pu)O 2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H 2 . Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructure). (author)

Compared performances of ENDF/B-VI and JEF-2.2 for MOX core physics

International Nuclear Information System (INIS)

Finck, P. J.

1998-01-01

The US is currently evaluating the use of MOX fuel in commercial LWR's for reducing weapons grade Pu stockpiles. The design and licensing processes will require that the validity of the nuclear data libraries and codes used in the effort be demonstrated. Unfortunately, there are only a very limited number of relatively old and non representative integral experiments' freely available to the US programs. This lack of adequate experimental data can be partially remediated by comparing the results of well validated European codes with the results of candidate US codes. The demonstration can actually be divided in two components: a code to code (Monte Carlo) comparison can easily demonstrate the validity and limits of the proposed algorithms; and the performances of nuclear data libraries should be compared, major trends should be observed, and their origins should be explained in terms of differences in evaluated nuclear data; In this paper, we have compared the performances of the JEF-2.2 and ENDF/B-VI.4 libraries for a series of benchmarks for k eff , void worth, and pin power distributions. Note that JEF-2.2 has been extensively validated for MOX applications

An immunocapture/scintillation proximity analysis of G alpha q/11 activation by native serotonin (5-HT)2A receptors in rat cortex: blockade by clozapine and mirtazapine.

Science.gov (United States)

Mannoury La Cour, C; Chaput, C; Touzard, M; Millan, M J

2009-02-01

Though transduction mechanisms recruited by heterologously expressed 5-HT(2A) receptors have been extensively studied, their interaction with specific subtypes of G-protein remains to be directly evaluated in cerebral tissue. Herein, as shown by an immunocapture/scintillation proximity analysis, 5-HT, the prototypical 5-HT(2A) agonist, DOI, and Ro60,0175 all enhanced [(35)S]GTPgammaS binding to G alpha q/11 in rat cortex with pEC(50) values of 6.22, 7.24 and 6.35, respectively. No activation of G o or G s/olf was seen at equivalent concentrations of DOI. Stimulation of G alpha q/11 by 5-HT (30 microM) and DOI (30 microM) was abolished by the selective 5-HT(2A) vs. 5-HT(2C)/5-HT(2B) antagonists, ketanserin (pK(B) values of 9.11 and 8.88, respectively) and MDL100,907 (9.82 and 9.68). By contrast, 5-HT-induced [(35)S]GTPgammaS binding to G alpha q/11 was only weakly inhibited by the preferential 5-HT(2C) receptor antagonists, RS102,221 (6.94) and SB242,084 (7.39), and the preferential 5-HT(2B) receptor antagonist, LY266,097 (6.66). The antipsychotic, clozapine, which had marked affinity for 5-HT(2A) receptors, blocked the recruitment of G alpha q/11 by 5-HT and DOI with pK(B) values of 8.54 and 8.14, respectively. Its actions were mimicked by the "atypical" antidepressant and 5-HT(2A) receptor antagonist, mirtazapine, which likewise blocked 5-HT and DOI-induced G alpha q/11 protein activation with pK(B) values of 7.90 and 7.76, respectively. In conclusion, by use of an immunocapture/scintillation proximity strategy, this study shows that native 5-HT(2A) receptors in rat frontal cortex specifically recruit G alpha q/11 and that this action is blocked by clozapine and mirtazapine. Quantification of 5-HT(2A) receptor-mediated G alpha q/11 activation in frontal cortex should prove instructive in characterizing the actions of diverse classes of psychotropic agent. 2008 Wiley-Liss, Inc.

Control of nuclear material hold-up: The key factors for design and operation of MOX fuel fabrication plants in Europe

International Nuclear Information System (INIS)

Beaman, M.; Beckers, J.; Boella, M.

2001-01-01

Full text: Some protagonists of the nuclear industry suggest that MOX fuel fabrication plants are awash with nuclear materials which cannot be adequately safeguarded and that materials 'stuck in the plant' could conceal clandestine diversion of plutonium. In Europe the real situation is quite different: nuclear operators have gone to considerable efforts to deploy effective systems for safety, security, quality and nuclear materials control and accountancy which provide detailed information. The safeguards authorities use this information as part of the safeguards measures enabling them to give safeguards assurances for MOX fuel fabrication plants. This paper focuses on the issue of hold-up: definition of the hold-up and of the so-called 'hidden inventory'; measures implemented by the plant operators, from design to day to day operations, for minimising hold-up and 'hidden inventory'; plant operators' actions to manage the hold-up during production activities but also at PIT/PIV time; monitoring and management of the 'hidden inventory'; measures implemented by the safeguards authorities and inspectorate for verification and control of both hold-up and 'hidden inventory'. The examples of the different plant specific experiences related in this paper reveal the extensive experience gained in european MOX fuel fabrication plants by the plant operators and the safeguards authorities for the minimising and the control of both hold-up and 'hidden inventory'. MOX fuel has been fabricated in Europe, with an actual combined capacity of 2501. HM/year subject, without any discrimination, to EURATOM Safeguards, for more than 30 years and the total output is, to date, some 1000 t.HM. (author)

dC-dG alternating oligonucleotides: Thermodynamic and kinetic aspects of the B-Z transformation

International Nuclear Information System (INIS)

Manzini, G.; Xodo, L.E.; Quadrifoglio, F.; van Boom, J.H.; van der Marel, G.A.

1987-01-01

The alternating cytosine-guanine oligodeoxyribonucleotides (dCdG)n, (dGdC)n, (dCdG)ndC (n = 3,4), (dGdC)7 and dG(dCdG)3 have been studied by UV and CD spectroscopy at different temperatures and NaCl concentrations. The analysis of the melting data, assuming an all-or-none model, reveals that in the B-conformation the 5'G/C3' stacking interactions are enthalpically favoured with respect to the 5'C/G3' one. The CD investigation of the B-Z equilibrium shows that the Z-conformation is enthalpically stabilized, while the B-conformation is entropically favoured, in the range of NaCl concentration considered (1 to 5 M). The kinetic data for the B-Z transformation, obtained with a salt-jump technique for the hexamer (dCdG)3, support a mechanism by which the Watson-Crick hydrogen bonds are broken before the bases flip over separately and eventually stack, reforming the H-bonds, in the new helix

Optimization of MOX fuel cycles in pebble bed HTGR

International Nuclear Information System (INIS)

Wei Jinfeng; Li Fu; Sun Yuliang

2013-01-01

Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

Transportation and packaging issues involving the disposition of surplus plutonium as MOX fuel in commercial LWRs

International Nuclear Information System (INIS)

Ludwig, S.B.; Welch, D.E.; Best, R.E.; Schmid, S.P.

1997-08-01

This report provides a view of anticipated transportation, packaging, and facility handling operations that are expected to occur at mixed-oxide (MOX) fuel fabrication and commercial reactor facilities. This information is intended for use by prospective contractors to the U.S. Department of Energy (DOE) who plan to submit proposals to DOE to manufacture and irradiate MOX fuel assemblies in domestic commercial light-water reactors. The report provides data to prospective consortia regarding packaging and pickup of MOX nuclear fuel assemblies at a MOX fuel manufacturing plant and transport and delivery of the MOX assemblies to nuclear power plants. The report also identifies areas where data are incomplete either because of the status of development or lack of sufficient information and specificity regarding the nuclear power plant(s) where deliveries will take place

Radial power density distribution of MOX fuel rods in the IFA-651

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

Transport of fresh MOX fuel assemblies for the Monju initial core

International Nuclear Information System (INIS)

Kurakami, J.; Ouchi, Y.; Usami, M.

1997-01-01

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

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2000-11-01

A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

International Nuclear Information System (INIS)

Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

2009-01-01

The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

Extension of the Si:C Stressor Thickness by Using Multiple ClusterCarbon Species

International Nuclear Information System (INIS)

Sekar, Karuppanan; Krull, Wade

2011-01-01

ClusterCarbon implantation is now well established as an attractive alternative for producing stress in advanced NMOS devices. ClusterCarbon has the advantage over monomer carbon implant in it's self-amorphization feature, eliminating the need for PAI implantation while producing highly substitutional carbon incorporation. To date, the limitation of this approach has been the high energy limit, due to the extraction limit of the available production tools for the preferred carbon species, which has been the C7Hx molecule. It is noted that the C7 species is produced by the breakup of the parent C14H14 molecule in the ion source. It is further noted that the preferred method of producing the Si:C stress layer is a multiple implant sequence with ClusterCarbon implants at various energies and doses designed to produce a carbon profile which is constant in-depth. The stressor thickness limit using C7 is known to be about 40 nm, which is less than the stressor thickness used in the conventional SiGe process for PMOS. In this work, it is shown that utilizing the C5 molecule which is also available from the breakup of C14H14 enables the stressor layer thickness to be extended to at least 60 nm, which is consistent with the conventional SiGe process. It will be shown that one additional C5 implant, performed after a standard C7 multiple implant sequence, can produce the extension of the stressor thickness while maintaining the flat depth profile. A detailed process characterization will be shown for this new process sequence.

Optical identifications of radio sources in the 5C 7 survey

International Nuclear Information System (INIS)

Perryman, M.A.C.

1979-01-01

An identification procedure developed for the deep radio survey 5C 6 has been refined and applied to the 5C 7 survey. Positions and finding charts are presented for candidate identifications from deep plates taken with the Palomar 48-inch Schmidt telescope. The identification statistics are in good agreement with the 5C 6 results, the accurate radio positions obtained at 1407 MHz defining a reasonably reliable and complete sample of associations with an identification rate of about 40 per cent. At 408 MHz the positional uncertainties are larger and the identifications are thus of lower reliability; the identification rate is about 20 per cent. The results are in good agreement with the assumptions that the optical identifications are coincident with the radio centroids, and that the identifications are not preferentially associated with faint clusters. (author)

Novel technique for manipulating MOX fuel particles using radiation pressure of a laser light

International Nuclear Information System (INIS)

Omori, R.; Suzuki, A.

2001-01-01

We proposed two principles based on the laser manipulation technique for collecting MOX fuel particles floating in air. While Principle A was based on the acceleration of the MOX particles due to the radiation pressure of a visible laser light, Principle B was based on the gradient forces exerted on the particles when an infrared laser light was incident. Principle A was experimentally verified using MnO 2 particles. Numerical results also showed the possibility of collecting MOX fuel particles based on both the principles. (authors)

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

International Nuclear Information System (INIS)

Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

2011-01-01

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

International Nuclear Information System (INIS)

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

2011-01-01

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

Toward full MOX core design

International Nuclear Information System (INIS)

Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

1999-01-01

This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

The 5-HT₁A receptor C(1019)G polymorphism influences the intravaginal ejaculation latency time in Dutch Caucasian men with lifelong premature ejaculation.

Science.gov (United States)

Janssen, Paddy K C; van Schaik, R; Zwinderman, Aeilko H; Olivier, Berend; Waldinger, Marcel D

2014-06-01

Lifelong premature ejaculation (LPE) is characterized by persistent intravaginal ejaculation latency times (IELTs) of less than 1 min, and has been postulated as a neurobiological dysfunction related to diminished serotonergic neurotransmission with 5-HT₁A receptor hyperfunction and 5-HT₂C hypofunction. To investigate the relationship between 5-HT₁A receptor gene (HTR₁A)-C(1019)G promoter polymorphism and IELT in men with LPE. This polymorphism is known to increase 5-HT1A receptor expression. A prospective study was conducted in 54 Dutch Caucasian men with LPE. Baseline IELT during coitus was assessed by stopwatch over a 1-month period. All men were genotyped for HTR₁A gene polymorphism. Allele frequencies and genotypes of C and G variants of HTR₁A polymorphism were determined. Association between CC, CG, and GG genotypes and the IELT in men with LPE were investigated. IELT measured by stopwatch, HTR₁A polymorphism. In this cohort of men with LPE, the geometric mean IELT was 23.8 s. Of the 54 men, the CC, CG and GG genotype frequency for the C(1019)G polymorphism of the 5-HT₁A gene was 33%, 43% and 24%, respectively. The geometric mean IELT for the CC, CG and GG genotypes were 14.5, 27.7 and 36.0 s, respectively (p=0.019). Compared to GG and CG genotypes, men with CC genotype had a 250% and 190% shorter ejaculation time, respectively. HTR₁A gene polymorphism is associated with the IELT in men with LPE. Men with CC genotype have shorter IELTs than men with GG and CG genotypes. Copyright © 2014 Elsevier Inc. All rights reserved.

Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities

International Nuclear Information System (INIS)

Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie

2005-01-01

It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-06-01

The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

MOX fuel: a contribution to disarmament. U.S. utilities' response to DOE's plutonium disposition decision

International Nuclear Information System (INIS)

Wallace, M.

1997-01-01

The author is chairman of the Nuclear Energy Institute Plutonium Disposition Working Group, which includes 11 nuclear utilities, including Ontario Hydro, and all the European fabricators of mixed oxide (MOX) fuel. A feasibility study is going on, to see if Russian or other weapons grade plutonium made into MOX fuel can be used in US, Canadian, or other power reactors. The US nuclear power industry is going through a period of change, and its primary responsibility must be the safe, reliable and economic operation of its plants. There is no current US MOX capacity, but the Europeans have have manufactured and burned over 400 tons of MOX fuel since 1963. Canada may be involved, initially through a pilot-scale experiment in NRU reactor

Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

Energy Technology Data Exchange (ETDEWEB)

Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

2014-02-15

Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

Determination of thermo-physical properties of (Th,Ce)O2 MOX using classical molecular dynamics simulations

International Nuclear Information System (INIS)

Ghosh, P.S.; Arya, A.; Dey, G.K.; Somayajulu, P.S.

2015-01-01

This paper calculates lattice thermal expansion (LTE) and thermal conductivity (TC) of Th 1-x Ce x O 2 (x=0.0, 0.0625 and 0.125) MOX using classical molecular dynamic simulations. The potential parameters of Coulomb-Buckingham function for Th-O, Ce-O and O-O were determined by fitting experimentally available LTE data for pure ThO 2 and CeO 2 . The calculated linear thermal expansion coefficients in the temperature range 300-1500K for ThO 2 , CeO 2 , Th 93.75 Ce 6.25 O 2 and Th 87.5 Ce 12.5 O 2 are 10.61, 13.08, 10.78 and 10.93x10 -6 K -1 , respectively. The MD calculated LTE values of ThO 2 and (Th,Ce)O 2 MOX are slightly higher than the experimentally determined values. The MD calculated TC values of ThO 2 and (Th,Ce)O 2 MOX in the high temperature range (600 to 1200 K) our results accords very well with the experimental measurements and at the low temperature range (300-500 K) our results are slightly different from some experimental results as the difference comes from our presumption that the dominant mechanism for phonon scattering is the Umklapp process. (author)

Hot vacuum outgassing to ensure low hydrogen content in MOX fuel pellets for thermal reactors

International Nuclear Information System (INIS)

Majumdar, S.; Nair, M.R.; Kumar, Arun

1983-01-01

Hot vacuum outgassing treatment to ensure low hydrogen content in Mixed Oxide Fuel (MOX) pellets for thermal reactors has been described. Hypostoichiometric sintered MOX pellets retain more hydrogen than UO 2 pellets. The hydrogen content further increases with the addition of admixed lubricant and pore formers. However, low hydrogen content in the MOX pellets can be ensured by a hot vacuum outgassing treatment at a temperature between 773K to 823K for 2 hrs. (author)

Evaluation of remaining behavior of halogen on the fabrication of MOX pellet containing Am

International Nuclear Information System (INIS)

Ozaki, Yoko; Osaka, Masahiko; Obayashi, Hiroshi; Tanaka, Kenya

2004-11-01

It is important to limit the content of halogen elements, namely fluorine and chlorine that are sources of making cladding material corrode, in nuclear fuel from the viewpoint of quality assurance. The halogen content should be more carefully limited in the MOX fuel containing Americium (Am-MOX), which is fabricated in the Alpha-Gamma Facility (AGF) for irradiation testing to be conducted in the experimental fast reactor JOYO, because fluorine may remain in the sintered pellets owing to a formation of AmF 3 known to have a low vapor pressure and may exceeds the limit of 25 ppm. In this study, a series of experimental determination of halogen element in Am-MOX were performed by a combination method of pyrolysis and ion-chromatography for the purpose of an evaluation of behavior of remaining halogen through the sintering process. Oxygen potential, temperature and time were changed as experimental parameters and their effects on the remaining behavior of halogen were examined. It was confirmed that good pellets, which contained small amount of halogen, could be obtained by the sintering for 3 hour at 1700degC in the oxygen potential range from -520 to -390 kJ/mol. In order to analysis of fluorine chemical form in green pellet, thermal analysis was performed. AmF 3 and PuF 3 have been confirmed to remain in the green pellet. (author)

Cooccurrence of Multiple AmpC β-Lactamases in Escherichia coli, Klebsiella pneumoniae, and Proteus mirabilis in Tunisia.

Science.gov (United States)

Chérif, Thouraya; Saidani, Mabrouka; Decré, Dominique; Boutiba-Ben Boubaker, Ilhem; Arlet, Guillaume

2016-01-01

Over a period of 40 months, plasmid-mediated AmpC β-lactamases were detected in Tunis, Tunisia, in 78 isolates (0.59%) of Escherichia coli, Klebsiella pneumoniae, and Proteus mirabilis. In 67 isolates, only one ampC gene was detected, i.e., blaCMY-2-type (n = 33), blaACC (n = 23), blaDHA (n = 6) or blaEBC (n = 5). Multiple ampC genes were detected in 11 isolates, with the following distribution: blaMOX-2, blaFOX-3, and blaCMY-4/16 (n = 6), blaFOX-3 and blaMOX-2 (n = 3), and blaCMY-4 and blaMOX-2 (n = 2). A great variety of plasmids carrying these genes was found, independently of the species and the bla gene. If the genetic context of blaCMY-2-type is variable, that of blaMOX-2, reported in part previously, is unique and that of blaFOX-3 is unique and new. Copyright © 2015, American Society for Microbiology. All Rights Reserved.

A Bridging Water Anchors the Tethered 5-(3-Aminopropyl)-2′-deoxyuridine Amine in the DNA Major Groove Proximate to the N+2 C·G Base Pair: Implications for Formation of Interstrand 5′-GNC-3′ Cross-Links by Nitrogen Mustards‡

Science.gov (United States)

Wang, Feng; Li, Feng; Ganguly, Manjori; Marky, Luis A.; Gold, Barry; Egli, Martin; Stone, Michael P.

2009-01-01

Site-specific insertion of 5-(3-aminopropyl)-2′-deoxyuridine (Z3dU) and 7-deaza-dG into the Dickerson-Drew dodecamers 5′-d(C1G2C3G4A5A6T7T8C9Z10C11G12)-3′ · 5′-d(C13G14C15G16A17A18T19T20-C21Z22C23G24)-3′ (named DDDZ10) and 5′-d(C1G2C3G4A5A6T7X8C9Z10C11G12)-3′ · 5′-d(C13G14C15G16A17A18-T19X20C21Z22C23G24)-3′ (named DDD2+Z10)(X = Z3dU; Z = 7-deaza-dG) suggests a mechanism underlying the formation of interstrand N+2 DNA cross-links by nitrogen mustards, e.g., melphalan and mechlorethamine. Analysis of the DDD2+Z10 duplex reveals that the tethered cations at base pairs A5 · X20 and X8 · A17 extend within the major groove in the 3′-direction, toward conserved Mg2+ binding sites located adjacent to N+2 base pairs C3 · Z22 and Z10 · C15. Bridging waters located between the tethered amines and either Z10 or Z22 O6 stabilize the tethered cations and allow interactions with the N + 2 base pairs without DNA bending. Incorporation of 7-deaza-dG into the DDD2+Z10 duplex weakens but does not eliminate electrostatic interactions between tethered amines and Z10 O6 and Z22 O6. The results suggest a mechanism by which tethered N7-dG aziridinium ions, the active species involved in formation of interstrand 5′-GNC-3′ cross-links by nitrogen mustards, modify the electrostatics of the major groove and position the aziridinium ions proximate to the major groove edge of the N+2 C · G base pair, facilitating interstrand cross-linking. PMID:18549246

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

Cytotoxic effects of the synthetic oestrogens and androgens on Balb/c 3T3 and HepG2 cells

Directory of Open Access Journals (Sweden)

Minta Maria

2014-12-01

Full Text Available The aim of the study was to test and compare the cytotoxic potential of two synthetic oestrogens: diethylstilboestrol (DES and ethinyloestradiol (EE2 and two androgens: testosterone propionate (TP and trenbolone (TREN on two cell lines. The fibroblast cell line Balb/c 3T3 and the hepatoma cell line HepG2 were selected. To get more insight into the mode of toxic action, four methods were used, which evaluated different biochemical endpoints: mitochondrial activity (3-(4,5-dimethylthiazol-2-yl- 2,5-diphenyltetrazolium bromide reduction assay, lysosomal activity (neutral red uptake assay, total protein content, and lactate dehydrogenase release. Cytotoxicity was assessed after 24, 48, and 72 h exposure to eight concentrations ranging from 0.78 to 100 μg/mL. Concentration- and time- dependent effects were observed. Depending on the line and assay used, half maximal effective concentration after 72 h (EC50-72h values ranged as follows: DES 1-13.7 μg/mL (Balb/c 3T3 and 3.7-5.2 μg/mL (HepG2; EE2 2.1-14.3 μg/mL (Balb/c 3T3 and 1.8-7.8 μg/mL (HepG2; TP-14.9-17.5 μg/mL (Balb/c 3T3, and 63.9- 100 μg/mL (HepG2; and TREN 11.3-31.4 μg/mL (Balb/c 3T3 and 12.5-59.4 μg/mL (HepG2. The results revealed that oestrogens were more toxic than androgens and the most affected endpoint was mitochondrial activity. In contrast to oestrogens, for which EC50-72h values were similar in both lines and by all assays used, Balb/c 3T3 cells were more sensitive than HepG2 cells to TP.

Downregulation of 5-HT7 Serotonin Receptors by the Atypical Antipsychotics Clozapine and Olanzapine. Role of Motifs in the C-Terminal Domain and Interaction with GASP-1

DEFF Research Database (Denmark)

Manfra, Ornella; Van Craenenbroeck, Kathleen; Skieterska, Kamila

2015-01-01

have previously found that the atypical antipsychotics clozapine and olanzapine inhibited G protein activation and, surprisingly, induced both internalization and lysosomal degradation of 5-HT7 receptors. Here, we aimed to determine the mechanism of clozapine- and olanzapine-mediated degradation of 5......-HT7 receptors. In the C-terminus of the 5-HT7 receptor, we identified two YXXΦ motifs, LR residues, and a palmitoylated cysteine anchor as potential sites involved in receptor trafficking to lysosomes followed by receptor degradation. Mutating either of these sites inhibited clozapine- and olanzapine...... of clozapine or olanzapine to the 5-HT7 receptor leads to antagonist-mediated lysosomal degradation by exposing key residues in the C-terminal tail that interact with GASP-1....

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

International Nuclear Information System (INIS)

O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

1998-08-01

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

Pressure-induced structural and semiconductor-semiconductor transitions in C o0.5M g0.5C r2O4

Science.gov (United States)

Rahman, S.; Saqib, Hajra; Zhang, Jinbo; Errandonea, D.; Menéndez, C.; Cazorla, C.; Samanta, Sudeshna; Li, Xiaodong; Lu, Junling; Wang, Lin

2018-05-01

The effect of pressure on the structural, vibrational, and electronic properties of Mg-doped Cr bearing spinel C o0.5M g0.5C r2O4 was studied up to 55 GPa at room-temperature using x-ray diffraction, Raman spectroscopy, electrical transport measurements, and ab initio calculations. We found that the ambient-pressure phase is cubic (spinel-type, F d 3 ¯m ) and underwent a pressure-induced structural transition to a tetragonal phase (space group I 4 ¯m 2 ) above 28 GPa. The ab initio calculation confirmed this first-order phase transition. The resistivity of the sample decreased at low pressures with the existence of a low-pressure (LP) phase and started to increase with the emergence of a high-pressure (HP) phase. The temperature dependent resistivity experiments at different pressures illustrated the wide band gap semiconducting nature of both the LP and HP phases with different activation energies, suggesting a semiconductor-semiconductor transition at HP. No evidence of chemical decomposition or a semiconductor-metal transition was observed in our studies.

Aqueous poly(amidoamine) dendrimer G3 and G4 generations with several interior cores at pHs 5 and 7: a molecular dynamics simulation study.

Science.gov (United States)

Kavyani, Sajjad; Amjad-Iranagh, Sepideh; Modarress, Hamid

2014-03-27

Poly(amidoamine) (PAMAM) dendrimers play an important role in drug delivery systems, because the dendrimers are susceptible to gain unique features with modification of their structure such as changing their terminals or improving their interior core. To investigate the core improvement and the effect of core nature on PAMAM dendrimers, we studied two generations G3 and G4 PAMAM dendrimers with the interior cores of commonly used ethylendiamine (EDA), 1,5-diaminohexane (DAH), and bis(3-aminopropyl) ether (BAPE) solvated in water, as an aqueous dendrimer system, by using molecular dynamics simulation and applying a coarse-grained (CG) dendrimer force field. To consider the electrostatic interactions, the simulations were performed at two protonation states, pHs 5 and 7. The results indicated that the core improvement of PAMAM dendrimers with DAH produces the largest size for G3 and G4 dendrimers at both pHs 5 and 7. The increase in the size was also observed for BAPE core but it was not so significant as that for DAH core. By considering the internal structure of dendrimers, it was found that PAMAM dendrimer shell with DAH core had more cavities than with BAPE core at both pHs 5 and 7. Also the moment of inertia calculations showed that the generation G3 is more open-shaped and has higher structural asymmetry than the generation G4. Possessing these properties by G3, specially due to its structural asymmetry, make penetration of water beads into the dendrimer feasible. But for higher generation G4 with its relatively structural symmetry, the encapsulation efficiency for water molecules can be enhanced by changing its core to DAH or BAPE. It is also observed that for the higher generation G4 the effect of core modification is more profound than G3 because the core modification promotes the structural asymmetry development of G4 more significantly. Comparing the number of water beads that penetrate into the PAMAM dendrimers for EDA, DAH, and BAPE cores indicates a

International symposium on MOX fuel cycle technologies for medium and long-term deployment. Book of extended synopses

International Nuclear Information System (INIS)

1999-05-01

The purpose of the Symposium was to provide a forum to exchange information on MOX fuel cycle technologies with focus on how past experience is being or can be used to progress further, either for facing more demanding fabrication and utilization conditions or for extending into new processing or utilization domains. Presented papers covered the following topics: Current status and prospects concerning plutonium management and MOX fuel utilization; MOX fuel fabrication technology and quality control; Fuel design, performance and testing; In-core fuel management and advanced fuel cycle options; Safety analysis, licensing and safeguards; Transportation and management of irradiated MOX fuel

Plasminogen Activator Inhibitor-1 (PAI-1) gene 4G/5G alleles frequency distribution in the Lebanese population.

Science.gov (United States)

Shammaa, Dina M R; Sabbagh, Amira S; Taher, Ali T; Zaatari, Ghazi S; Mahfouz, Rami A R

2008-09-01

Plasminogen activator inhibitor-1 (PAI-1) is an inhibitor of fibrinolysis. Increased plasma PAI-1 levels play an essential role in the pathogenesis of cardiovascular risk and other diseases associated with thrombosis. The 4G/5G polymorphism of the PAI-1 promoter region has been extensively studied in different populations. We studied 160 healthy unrelated Lebanese individuals using a reverse hybridization PCR assay to detect the 5G/5G, 4G/5G and, 4G/4G genotypes of the PAI-1 gene and the frequencies of the 4G and 5G alleles. We found that 4G/5G genotype was the most prevalent (45.6%) followed by 5G/5G (36.9%) and 4G/4G (17.5%). The frequencies of the 4G and 5G alleles were calculated to be 0.403 and 0.597, respectively. Compared to other ethnic communities, the Lebanese population was found to harbour a relatively high prevalence of the rare 4G allele. This, in turn, may predispose this population to develop cardiovascular diseases and other thrombotic clinical conditions. This study aids to enhance our understanding of the genetic features of the Lebanese population.

Thorium utilization as a Pu-burner: proposal of Plutonium-Thorium Mixed Oxide (PT-MOX) Project

International Nuclear Information System (INIS)

Aizawa, Otohiko

2000-01-01

In this paper, a Pu-Th mixed oxide (PT-MOX) project is proposed for a thorium utilization and a plutonium burning. None of plutonium can be newly produced from PT-MOX fuel, and the plutonium mass of about 1 ton can be consumed with one reactor (total heavy metal assumed: 100 tons) for 1 year. In order to consume plutonium produced from usual Light Water Reactor, it should be better to operate one PT-MOX reactor for three to five Light Water Reactors. (author)

Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

Energy Technology Data Exchange (ETDEWEB)

Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-05-01

Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-08-01

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

Image analysis: a tool characterising and modelling the microstructure of the MOX fuel

International Nuclear Information System (INIS)

Charollais, F.

1997-01-01

The MOX nuclear fuel, made up of about 3 to 10 % of plutonium oxide mixed with uranium oxide, is elaborated by an original manufacturing method (MIMAS process). The MOX pellets feature a singular and complex microstructure, including enriched plutonium zones dispersed in a low plutonium content matrix. Their properties as well as their performances levels are strongly linked with this microstructure. Tools, found in the literature, allowing to quantify with relevant parameters the microstructural images from different analytical equipment (optical microscopy, electron probe micro-analyser and autoradiography) have been adapted and used in order to characterize these nuclear fuels. Taking into account the heterogeneity of the MOX microstructure, we turn our's attention, at the beginning of this study, to the analysis conditions: choice of the magnification, sampling and statistical analysis of the measurements. An improvement of the ceramographic preparation of the samples, required for an automatic image analysis (of the granular structure), has been realised by thermal etching under oxidizing gas. This method enables the strong content plutonium zones to be revealed distinctly. The first part of the study concerns the characterization of the three-dimensional structure of uranium oxide and MOX fuels by average variables using the principles of mathematical morphology and stereology. The second part introduces probabilistic models, in particular the Boolean scheme, in order to improve and complete the three-dimensional characterization of the MOX fuel and more specifically the enriched plutonium islands dispersion in the pellet. [fr

Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

Energy Technology Data Exchange (ETDEWEB)

Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

2013-09-15

Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

Buildup of radioxenon isotopes in MOX-assemblies

Energy Technology Data Exchange (ETDEWEB)

Gniffke, Thomas; Kirchner, Gerald [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research, Hamburg (Germany)

2015-07-01

Radioxenon is the main tracer for detection of nuclear tests conducted underground under the verification regime of the Comprehensive Nuclear Test Ban Treaty (CTBT). Since radioxenon is emitted by civilian sources too, like commercial nuclear reactors, source discrimination is still an important issue. Inventory calculations are necessary to predict which xenon isotopic ratios are built up in a reactor and how they differ from those generated by a nuclear explosion. The screening line actually used by the CTBT Organization for source discrimination is based on calculations for uranium fuel of various enrichments used in pressurized water reactors (PWRs). The usage of different fuel, especially mixed U/Pu oxide (MOX) assemblies with reprocessed plutonium, may alter the radioxenon signature of civilian reactors. In this talk, calculations of the radioxenon buildup in a MOX-assembly used in a commercial PWR are presented. Implications for the CTBT verification regimes are discussed and open questions are addressed.

Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

2007-01-01

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)

Molecular Characterization of Echinococcus granulosus Cysts in North Indian Patients: Identification of G1, G3, G5 and G6 Genotypes

Science.gov (United States)

Sharma, Monika; Sehgal, Rakesh; Fomda, Bashir Ahmad; Malhotra, Anil; Malla, Nancy

2013-01-01

Background Cystic echinococcosis (CE) caused by the Echinococcus granulosus, is a major public health problem worldwide, including India. The different genotypes of E. granulosus responsible for human hydatidosis have been reported from endemic areas throughout the world. However, the genetic characterization of E. granulosus infecting the human population in India is lacking. The aim of study was to ascertain the genotype(s) of the parasite responsible for human hydatidosis in North India. Methodology/Principal Findings To study the transmission patterns of E. granulosus, genotypic analysis was performed on hydatid cysts obtained from 32 cystic echinococcosis (CE) patients residing in 7 different states of North India. Mitochondrial cytochrome c oxidase subunit1 (cox1) sequencing was done for molecular identification of the isolates. Most of the CE patients (30/32) were found to be infected with hydatid cyst of either G3 (53.1%) or G1 (40.62%) genotype and one each of G5 (cattle strain) and G6 (camel strain) genotype. Conclusions/Significance These findings demonstrate the zoonotic potential of G1 (sheep strain) and G3 (buffalo strain) genotypes of E. granulosus as these emerged as predominant genotypes infecting the humans in India. In addition to this, the present study reports the first human CE case infected with G5 genotype (cattle strain) in an Asian country and presence of G6 genotype (camel strain) in India. The results may have important implications in the planning of control strategies for human hydatidosis. PMID:23785531

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

2004-07-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

Decommissioning the Belgonucleaire Dessel MOX plant: presentation of the project and situation end august 2013

Energy Technology Data Exchange (ETDEWEB)

Cuchet, J.M. [TRACTEBEL ENGINEERING, Avenue Ariane, 7, B1200 Brussels (Belgium); Libon, H.; Verheyen, C. [BELGONUCLEAIRE S.A. / N.V. Europalaan, 20, B2480 Dessel (Belgium); Bily, J. [STUDSVIK GmbH, Karlsruher Strasse, 20, D75179 Pforzheim,(Germany); Boden, S. [SCK-CEN, Boeretang, 200, B2400 Mol (Belgium); Joffroy, F. [TECNUBEL N.V., Zandbergen, 1, B2480 Dessel (Belgium); Walthery, R. [BELGOPROCESS, Gravenstraat, 73, B2480 Dessel (Belgium)

2013-07-01

Belgonucleaire has been operating the Dessel MOX plant at an industrial scale between 1986 and 2006. During this period, 40 metric tons of plutonium (HM) have been processed into 90 reloads of MOX fuel for commercial light water reactors. The decision to stop the production in 2006 and to decommission the MOX plant was the result of the shrinkage of the MOX fuel market due to political and commercial factors. As a significant part of the decommissioning project of the Dessel MOX plant, about 170 medium-sized glove-boxes and about 1.200 metric tons of structure and equipment outside the glove-boxes are planned for dismantling. The license for the dismantling of the MOX plant was granted by Royal Decree in 2008 and the dismantling started in March 2009. The dismantling works are carried out by an integrated organization under leadership and responsibility of Belgonucleaire; this organization includes 3 main contractors, namely Tecnubel N.V., the THV ('Tijdelijke HandelsVereniging') Belgoprocess / SCK-CEN and Studsvik GmbH and Tractebel Engineering as project manager. In this paper, after having described the main characteristics of the project, the authors review the different organizational and technical options considered for the decommissioning of the glove-boxes; thereafter the main decision criteria (qualification of personnel and of processes, confinement, cutting techniques and radiation protection, safety aspects, alpha-bearing waste management) are analyzed as well. Finally the progress, the feedback and the lessons learned at the end of August 2013 are presented, giving the principal's and contractors point of view. (authors)

Photodissociation of C{sub 3}H{sub 5}Br and C{sub 4}H{sub 7}Br at 234 nm

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Kook; Paul, Dababrata; Hong, Ki Ryong; Cho, Ha Na; Kim, Tae Kyu [Pusan National University, Busan (Korea, Republic of); Lee, Kyoung Seok [Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

2012-01-15

The photodissociation dynamics of cyclopropyl bromide (C-3H{sub 5}Br) and cyclobutyl bromide (C{sub 4}H{sub 7}Br) at 234 nm was investigated. A two-dimensional photofragment ion-imaging technique coupled with a [2+1] resonance enhanced multiphoton ionization scheme was utilized to obtain speed and angular distributions of the nascent Br({sup 2}P{sub 3/2}) and Br*({sup 2}P{sub 1/2}) atoms. The recoil anisotropies for the Br and Br* channels were measured to be βBr = 0.92 ± 0.03 and βBr* = 1.52 ± 0.04 for C{sub 3}H{sub 5}Br and βBr = 1.10 ± 0.03 and βBr* = 1.49 ± 0.05 for C{sub 4}H{sub 7}Br. The relative quantum yield for Br was found to be ΦBr = 0.13 ± 0.03 and for C{sub 3}H{sub 5}Br and C{sub 4}H{sub 7}Br, respectively. The soft radical limit of the impulsive model adequately modeled the related energy partitioning. The nonadiabatic transition probability from the 3A' and 4A' potential energy surfaces was estimated and discussed.

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Sodium fast reactor: an asset for a PWR UOX/MOX fleet - 5327

International Nuclear Information System (INIS)

Tiphine, M.; Coquelet-Pascal, C.; Girieud, R.; Eschbach, R.; Chabert, C.; Grosman, R.

2015-01-01

Due to its low fissile content, Pu from spent MOX fuels is sometimes regarded as not recyclable in LWR. Based on the existing French nuclear infrastructure (La Hague reprocessing plant and MELOX MOX manufacturing plant), AREVA and CEA have evaluated the conditions of Pu multi recycling in a 100% LWR fleet. As France is currently supporting a Fast Reactor prototype project, scenario studies have also been conducted to evaluate the contribution of a 600 MWe SFR in the LWR fleet. These scenario studies consider a nuclear fleet composed of 8 PWR 900 MWe, with or without the contribution of a SFR, and aim at evaluating the following points: -) the feasibility of Pu multi-recycling in PWR; -) the impact on the spent fuels storage; -) the reduction of the stored separated Pu; -) the impact on waste management and final disposal. The studies have been conducted with the COSI6 code, developed by CEA Nuclear Energy Direction since 1985, that simulates the evolution over time of a nuclear power plants fleet and of its associated fuel cycle facilities and provides material flux and isotopic compositions at each point of the scenario. To multi-recycle Pu into LWR MOX and to ensure flexibility, different reprocessing strategies were evaluated by adjusting the reprocessing order, the choice of used fuel assemblies according to their burn-up and the UOX/MOX proportions. The improvement of the Pu fissile quality and the kinetic of Pu multi-recycling in SFR depending on the initial Pu quality were also evaluated and led to a reintroduction of Pu in PWR MOX after a single irradiation in SFR, still in dilution with Pu from UOX to maintain a sufficient fissile quality. (authors)

A comparative study of fission gas behaviour in UO2 and MOX fuels using the meteor fuel performance code

International Nuclear Information System (INIS)

Struzik, C.; Garcia, Ph.; Noirot, L.

2002-01-01

The paper reviews some of the fission-gas-related differences observed between MOX MIMAS AUC fuels and homogeneous UO 2 fuels. Under steady-state conditions, the apparently higher fractional release in MOX fuels is interpreted with the METEOR fuel performance code as a consequence of their lower thermal conductivity and the higher linear heat rates to which MOX fuel rods are subjected. Although more fundamental diffusion properties are needed, the apparently greater swelling of MOX fuel rods at higher linear heat rates can be ascribed to enhanced diffusion properties. (authors)

7 CFR 15b.27 - Extension education.

Science.gov (United States)

2010-01-01

... 7 Agriculture 1 2010-01-01 2010-01-01 false Extension education. 15b.27 Section 15b.27 Agriculture... Education § 15b.27 Extension education. (a) General. A recipient to which this subpart applies that provides extension education may not, on the basis of handicap, exclude qualified handicapped persons. A recipient...

Enantioselective route to 5-methyl- and 5,7-dimethyl-6,7-dihydro-5H-dibenz[c,e]azepine: secondary amines with switchable axial chirality.

Science.gov (United States)

Pira, Silvain L; Wallace, Timothy W; Graham, Jonathan P

2009-04-02

(-)-5-Methyl-6,7-dihydro-5H-dibenz[c,e]azepine 4, a new secondary amine featuring an axis-center stereochemical relay, was prepared enantioselectively from 2'-acetylbiphenyl-2-carboxylic acid, using (R)-2-phenylglycinol as an auxiliary for the control of both elements of chirality. The biaryl axis in 4 preferentially adopts the aS-configuration, with the methyl substituent pseudoequatorial, but conversion into the corresponding N-Boc derivative locks the axis into the aR-configuration, as predicted on the basis of molecular mechanics calculations.

International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

International Nuclear Information System (INIS)

Lovasic, Z.; Einziger, R.

2009-01-01

This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

Remarkable support effect on the reactivity of Pt/In2O3/MOx catalysts for methanol steam reforming

Science.gov (United States)

Liu, Xin; Men, Yong; Wang, Jinguo; He, Rong; Wang, Yuanqiang

2017-10-01

Effects of supports over Pt/In2O3/MOx catalysts with extremely low loading of Pt (1 wt%) and In2O3 loadings (3 wt%) are investigated for the hydrogen production of methanol steam reforming (MSR) in the temperature range of 250-400 °C. Under practical conditions without the pre-reduction, the 1Pt/3In2O3/CeO2 catalyst shows the highly efficient catalytic performance, achieving almost complete methanol conversion (98.7%) and very low CO selectivity of 2.6% at 325 °C. The supported Pt/In2O3 catalysts are characterized by means of Brunauer-Emmett-Teller (BET) surface area, X-ray diffraction (XRD), high-resolution transmission microscopy (HRTEM), temperature programmed reduction with hydrogen (H2-TPR), CO pulse chemisorption, temperature programmed desorption of methanol and water (CH3OH-TPD and H2O-TPD). These demonstrate that the nature of catalyst support of Pt/In2O3/MOx plays crucial roles in the Pt dispersion associated by the strong interaction among Pt, In2O3 and supporting materials and the surface redox properties at low temperature, and thus affects their capability to activate the reactants and determines the catalytic activity of methanol steam reforming. The superior 1Pt/3In2O3/CeO2 catalyst, exhibiting a remarkable reactivity and stability for 32 h on stream, demonstrates its potential for efficient hydrogen production of methanol steam reforming in mobile and de-centralized H2-fueled PEMFC systems.

A fission gas release model for MOX fuel and its verification

International Nuclear Information System (INIS)

Koo, Y.H.; Sohn, D.S.; Strijov, P.

2000-01-01

A fission gas release model for MOX fuel has been developed based on a model for UO 2 fuel. Using the concept of equivalent cell, the model considers the uneven distribution of Pu within the fuel matrix and a number of Pu-rich particles that could lead to a non-uniform fission rate and fission gas distribution across the fuel pellet. The model has been incorporated into a code, COSMOS, and some parametric studies were made to analyze the effect of the size and Pu content of Pu-rich agglomerates. The model was then applied to the experimental data obtained from the FIGARO program, which consisted of the base irradiation of MOX fuels in the BEZNAU-1 PWR and the subsequent irradiation of four refabricated fuel segments in the Halden reactor. The calculated gas releases show good agreement with the measured ones. In addition, the present analysis indicates that the microstructure of the MOX fuel used in the FIGARO program is such that it has produced little difference in terms of gas release compared with UO 2 fuel. (author)

Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

International Nuclear Information System (INIS)

Lippens, M.; Maldague, Th.; Basselier, J.; Boulanger, D.; Mertens, L.

2000-01-01

Part of the R and D work made at BELGONUCLEAIRE in the field of high burnup achievement with MOX fuel in commercial LWRs is made through lnternational Programmes. Special attention is given to the evolution with burnup of fuel neutronic characteristics and of in-reactor rod thermal-mechanical behaviour. Pu burning in MOX is characterized essentially by a drop of Pu 239 content. The other Pu isotopes have an almost unchanged concentration, due to internal breeding. The reactivity drop of MOX versus burnup is consequently much less pronounced than in UO 2 fuel. Concentration of minor actinides Am and Cm becomes significant with burnup increase. These nuclides start to play a role on total reactivity and in the helium production. The thermal-mechanical behaviour of MOX fuel rod is very similar to that of UO 2 . Some specificities are noticed. The better PCI resistance recognized to MOX fuel has recently been confirmed. Three PWR MOX segments pm-irradiated up to 58 GWd/tM were ramped at 100 W/cm.min respectively to 430-450-500 W/cm followed by a hold time of 24 hours. No segment failed. MOX and UO 2 fuels have different reactivities and operate thus at different powers. Moreover, radial distribution of power in MOX pellet is less depressed at high burnup than in UO 2 , leading to higher fuel central temperature for a same rating. The thermal conductivity of MOX fuel decreases with Pu content, typically 4% for 10% Pu. The combination of these three elements (power level, power profile, and conductivity) lead to larger FGR at high burnup compared to UO 2 . Helium production remains low compared to fission gas production (ratio < 0.2). As faster diffusing element, the helium fractional release is much higher than that of fission gas, leading to rod pressure increase comparable to the one resulting from fission gas. (author)

Nano Polyamidoamine-G7 (NPAMAM-G7 Dendrimer Synthesis and Assessment the Effect of its antibacterial on Escherichia Coli, Klebsiella Oxytoca, Pseudomonas Aeruginosa , Proteus Mirabilis and Staphylococcus Aureus from aqueous solution

Directory of Open Access Journals (Sweden)

mitra Gholami

2016-06-01

Full Text Available Background: Nano scale dendrimers are macromolecules synthetic which frequently used in medical and health field. This study was aimed to examine synthesis and the antibacterial effect of NanoPolyamidoamine-G7 (NPAMAM-G7 dendrimer on Escherichia Coli, Klebsiella Oxytoca, Pseudomonas Aeruginosa , Proteus Mirabilis and Staphylococcus Aureus from aqueous solution. Material and methods: In this experimental study, initially dilution of 103 CFU/ml were prepared from each strain of bacteria. Then different concentrations of dendrimer (0.025, 0.25, 2.5 and 25 µg/ml in the laboratory temperature (23-25 °C was added to water. In order to determine the efficiency of dendrimers in removal of bacteria, samples were taken at different times (0, 10, 20, 30, 40, 50 and 60 min and were cultured on nutrient agar medium. Samples were incubated for 24 hours at 37 °C and then number of colonies were counted. Results: Antibacterial properties of dendrimers in aqueous solution by increasing the dendrimer concentration and contact time is directly related. At a concentration of 25 μg/ml at 60 minutes all bacteria except Staphylococcus Aureus, and at 30 minutes, Escherichia Coli and Klebsiella Oxytoca bacteria for 100% excluded. The concentration of 2.5 μg/ml at 60 minutes of bacteria, Escherichia Coli, Klebsiella Oxytoca and Proteus Mirabilis are 100% excluded. All concentrations of dendrimers at different times were reduced bacteria in the PAMAM- G7 dendrimer effect on gram-negative bacteria, gram-positive bacteria was better. Conclusion: The NPAMAM-G7 dendrimer with end amine groups exhibited a positive impact on the removal of standard strains, gram-positive and gram-negative bacteria. Therefore, it is possible to use these nanodendrimers as antibacterial in the future.

Labeling and preliminary in vivo evaluation of the 5-HT7 receptor selective agonist [(11)C]E-55888

DEFF Research Database (Denmark)

Hansen, Hanne D; Andersen, Valdemar L; Lehel, Szabolcs

2015-01-01

E-55888 has been identified as a selective serotonin 7 (5-HT7) receptor agonist. In this study, we describe the synthesis, radiolabeling and in vivo evaluation of [(11)C]E-55888 as a radioligand for positron emission tomography (PET) imaging. [(11)C]E-55888 was obtained by N-methylation of an app...... neither be displaced by the structurally different 5-HT7 receptor ligand SB-269970 nor by self-block with unlabeled E-55888. Based on these data, [(11)C]E-55888 does not show promise as a PET radioligand for imaging the 5-HT7 receptor in vivo....

What is 5G? Emerging 5G Mobile Services and Network Requirements

Directory of Open Access Journals (Sweden)

Heejung Yu

2017-10-01

Full Text Available In this paper, emerging 5G mobile services are investigated and categorized from the perspective of not service providers, but end-users. The development of 5G mobile services is based on an intensive analysis of the global trends in mobile services. Additionally, several indispensable service requirements, essential for realizing service scenarios presented, are described. To illustrate the changes in societies and in daily life in the 5G era, five megatrends, including the explosion of mobile data traffic, the rapid increase in connected devices, everything on the cloud, hyper-realistic media for convergence services and knowledge as a service enabled by big-data analysis, are examined. Based on such trends, we classify the new 5G services into five categories in terms of the end-users’ experience as follows: immersive 5G services, intelligent 5G services, omnipresent 5G services, autonomous 5G services and public 5G services. Moreover, several 5G service scenarios in each service category are presented, and essential technical requirements for realizing the aforementioned 5G services are suggested, along with a competitiveness analysis on 5G services/devices/network industries and the current condition of 5G technologies.

A facility for creating Python extensions in C++

International Nuclear Information System (INIS)

Dubois, P F

1998-01-01

Python extensions are usually created by writing the glue that connects Python to the desired new functionality in the C language. While simple extensions do not require much effort, to do the job correctly with full error checking is tedious and prone to errors in reference counting and to memory leaks, especially when errors occur. The resulting program is difficult to read and maintain. By designing suitable C++ classes to wrap the Python C API, we are able to produce extensions that are correct and which clean up after themselves correctly when errors occur. This facility also integrates the C++ and Python exception facilities. This paper briefly describes our package for this purpose, named CXX. The emphasis is on our design choices and the way these contribute to the construction of accurate Python extensions. We also briefly relate the way CXX's facilities for sequence classes allow use of C++'s Standard Template Library (STL) algorithms on C++ sequences

Novel ternary g-C3N4/Ag3VO4/AgBr nanocomposites with excellent visible-light-driven photocatalytic performance for environmental applications

Science.gov (United States)

Barzegar, Javid; Habibi-Yangjeh, Aziz; Akhundi, Anise; Vadivel, S.

2018-04-01

Novel visible-light-induced photocatalysts were fabricated by integration of Ag3VO4 and AgBr semiconductors with graphitic carbon nitride (g-C3N4) through a facile refluxing method. The fabricated photocatalysts were extensively characterized by XRD, EDX, SEM, TEM, FT-IR, UV-vis DRS, BET, TGA, and PL instruments. The photocatalytic performance of these samples was studied by degradations of three dye contaminants under visible-light exposure. Among the ternary photocatalysts, the g-C3N4/Ag3VO4/AgBr (10%) nanocomposite displayed the maximum activity for RhB degradation with rate constant of 1366.6 × 10-4 min-1, which is 116, 7.23, and 38.5 times as high as those of the g-C3N4, g-C3N4/AgBr (10%), and g-C3N4/Ag3VO4 (30%) photocatalysts, respectively. The effects of synthesis time and calcination temperature were also investigated and discussed. Furthermore, according to the trapping experiments, it was found that superoxide anion radicals were the predominant reactive species in this system. Finally, the ternary photocatalyst displayed superlative activity in removal of the contaminants under visible-light exposure, displaying great potential of this ternary photocatalyst for environmental remediation, because of a facile synthesis route and outstanding photocatalytic performance.

First liquid chromatography method for the simultaneous determination of levofloxacin, pazufloxacin, gatifloxacin, moxifloxacin and trovafloxacin in human plasma.

Science.gov (United States)

Sousa, Joana; Alves, Gilberto; Campos, Gonçalo; Fortuna, Ana; Falcão, Amílcar

2013-07-01

For the first time a simple, selective and sensitive liquid chromatography method was developed and validated for the simultaneous determination of levofloxacin (LEV), pazufloxacin (PAZ), gatifloxacin (GAT), moxifloxacin (MOX) and trovafloxacin (TRO) in human plasma. Samples were pre-treated with acetonitrile for precipitation of plasma proteins followed by evaporation and reconstitution steps. Chromatographic separation of the analytes and norfloxacin, used as internal standard (IS), was performed under gradient elution on a LiChroCART(®) Purospher Star C18 column (55mm×4mm, 3μm). The mobile phase comprised a mixture of 0.1% aqueous formic acid adjusted to pH 3.0 with triethylamine, acetonitrile and methanol pumped at a flow rate of 1.0mL/min. The detector was set at excitation/emission wavelengths of 260/455nm. Calibration curves were linear (r(2)≥0.9923) in the ranges of 0.005-5μg/mL for GAT, 0.02-5μg/mL for LEV, PAZ and MOX and 0.04-5μg/mL for TRO. The intra and interday precision did not exceed 7.32% and the intra and interday accuracy ranged from -11.73 to 8.92%. The limits of quantification were established at 0.005μg/mL for GAT, 0.02μg/mL for LEV, PAZ and MOX and 0.04μg/mL for TRO. No endogenous or tested exogenous compounds were found to interfere at the retention times of the analytes and IS. Since the proposed method proved to be reliable for the quantitative determination of LEV, PAZ, GAT, MOX and TRO it may be a useful tool for routine analysis and to support clinical pharmacokinetic and toxicological studies involving these antibiotics. Copyright © 2013 Elsevier B.V. All rights reserved.

On the potential of OFDM enhancements as 5G waveforms

DEFF Research Database (Denmark)

Berardinelli, Gilberto; Pajukoski, Kari; Lähetkangas, Eeva

2014-01-01

The ideal radio waveform for an upcoming 5th Generation (5G) radio access technology should cope with a set of requirements such as limited complexity, good time/frequency localization and simple extension to multi-antenna technologies. This paper discusses the suitability of Orthogonal Frequency...... Division Multiplexing (OFDM) and its recently proposed enhancements as 5G waveforms, mainly focusing on their capability to cope with our requirements. Significant focus is given to the novel zero-tail paradigm, which allows boosting the OFDM flexibility while circumventing demerits such as poor spectral...

Simulation of facility operations and materials accounting for a combined reprocessing/MOX fuel fabrication facility

International Nuclear Information System (INIS)

Coulter, C.A.; Whiteson, R.; Zardecki, A.

1991-01-01

We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs

Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

International Nuclear Information System (INIS)

Wachs, G.W.

1997-11-01

The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

Molecular characterization of Echinococcus granulosus cysts in north Indian patients: identification of G1, G3, G5 and G6 genotypes.

Directory of Open Access Journals (Sweden)

Monika Sharma

Full Text Available BACKGROUND: Cystic echinococcosis (CE caused by the Echinococcus granulosus, is a major public health problem worldwide, including India. The different genotypes of E. granulosus responsible for human hydatidosis have been reported from endemic areas throughout the world. However, the genetic characterization of E. granulosus infecting the human population in India is lacking. The aim of study was to ascertain the genotype(s of the parasite responsible for human hydatidosis in North India. METHODOLOGY/PRINCIPAL FINDINGS: To study the transmission patterns of E. granulosus, genotypic analysis was performed on hydatid cysts obtained from 32 cystic echinococcosis (CE patients residing in 7 different states of North India. Mitochondrial cytochrome c oxidase subunit1 (cox1 sequencing was done for molecular identification of the isolates. Most of the CE patients (30/32 were found to be infected with hydatid cyst of either G3 (53.1% or G1 (40.62% genotype and one each of G5 (cattle strain and G6 (camel strain genotype. CONCLUSIONS/SIGNIFICANCE: These findings demonstrate the zoonotic potential of G1 (sheep strain and G3 (buffalo strain genotypes of E. granulosus as these emerged as predominant genotypes infecting the humans in India. In addition to this, the present study reports the first human CE case infected with G5 genotype (cattle strain in an Asian country and presence of G6 genotype (camel strain in India. The results may have important implications in the planning of control strategies for human hydatidosis.

Rapid Industrial Prototyping and SoC Design of 3G/4G Wireless Systems Using an HLS Methodology

Directory of Open Access Journals (Sweden)

Andres Takach

2006-07-01

Full Text Available Many very-high-complexity signal processing algorithms are required in future wireless systems, giving tremendous challenges to real-time implementations. In this paper, we present our industrial rapid prototyping experiences on 3G/4G wireless systems using advanced signal processing algorithms in MIMO-CDMA and MIMO-OFDM systems. Core system design issues are studied and advanced receiver algorithms suitable for implementation are proposed for synchronization, MIMO equalization, and detection. We then present VLSI-oriented complexity reduction schemes and demonstrate how to interact these high-complexity algorithms with an HLS-based methodology for extensive design space exploration. This is achieved by abstracting the main effort from hardware iterations to the algorithmic C/C++ fixed-point design. We also analyze the advantages and limitations of the methodology. Our industrial design experience demonstrates that it is possible to enable an extensive architectural analysis in a short-time frame using HLS methodology, which significantly shortens the time to market for wireless systems.

Rapid Industrial Prototyping and SoC Design of 3G/4G Wireless Systems Using an HLS Methodology

Directory of Open Access Journals (Sweden)

Cavallaro JosephR

2006-01-01

Full Text Available Many very-high-complexity signal processing algorithms are required in future wireless systems, giving tremendous challenges to real-time implementations. In this paper, we present our industrial rapid prototyping experiences on 3G/4G wireless systems using advanced signal processing algorithms in MIMO-CDMA and MIMO-OFDM systems. Core system design issues are studied and advanced receiver algorithms suitable for implementation are proposed for synchronization, MIMO equalization, and detection. We then present VLSI-oriented complexity reduction schemes and demonstrate how to interact these high-complexity algorithms with an HLS-based methodology for extensive design space exploration. This is achieved by abstracting the main effort from hardware iterations to the algorithmic C/C++ fixed-point design. We also analyze the advantages and limitations of the methodology. Our industrial design experience demonstrates that it is possible to enable an extensive architectural analysis in a short-time frame using HLS methodology, which significantly shortens the time to market for wireless systems.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

Energy Technology Data Exchange (ETDEWEB)

Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

2014-06-01

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Evaluation of the characteristics of uranium and plutonium Mixed Oxide (MOX) fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

MOX fuel irradiation test up to high burnup has been performed for five years. Irradiation test of MOX fuel having high plutonium content has also been performed from JFY 2007 and it still continues. A lot of irradiation data have been obtained through these tests. The activities done in JFY 2012 are mainly focused on Post Irradiation Examination (PIE) data analysis concerning thermal property change and fission gas release. In the former work thermal conductivity degradation due to burnup is examined and in the latter work the dependence of fission gas release mechanism on fuel pellet microstructure is examined. This report mainly covers the result of analysis. It is found that thermal conductivity degradation of MOX fuel due to burnup is less than that of UO{sub 2} fuel and that fission gas release mechanism of high enriched fissile zone (so called Pu spot) is much different from that of low enriched fissile zone (so called Matrix). (author)

A PCI failure in an experimental MOX fuel rod and its sensitivity analysis

International Nuclear Information System (INIS)

Marino, A.C.

2000-01-01

Within our interest in studying MOX fuel performance, the irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986 with six rods fabricated at the α Facility (CNEA, Argentina). These experiences were made in the HFR-Petten reactor, Holland. The goal of this experience was to study the fuel behaviour with respect to PMCI-SCC. An experiment for extended burnup was performed with the last two MOX rods. During the experiment the final test ramp was interrupted due to a failure in the rod. The post-irradiation examinations indicated that PCI-SCC was a mechanism likely to produce the failure. At the Argentine Atomic Energy Commission (CNEA) the BACO code was developed for the simulation of a fuel rod thermo-mechanical behaviour under stationary and transient power situations. BACO includes a probability analysis within its structure. In BACO the criterion for safe operation of the fuel is based on the maximum hoop stress being below a critical value at the cladding inner surface; this is related to susceptibility to stress corrosion cracking (SCC). The parameters of the MOX irradiation, the preparation of the experiments and post-irradiation analysis were sustained by the BACO code predictions. We present in this paper an overview of the different experiences performed with the MOX fuel rods and the main findings of the post-irradiation examinations. A BACO code description, a wide set of examples which sustain the BACO code validation, and a special calculation for BU15 experiment attained using the BACO code, including a probabilistic analysis of the influence of rod parameters on performance, are included. (author)

Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

Energy Technology Data Exchange (ETDEWEB)

Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

1996-12-31

The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

Synthesis and Characterization of Nanocrystalline Ni50Al50-xMox (X=0-5 Intermetallic Compound During Mechanical Alloying Process

Directory of Open Access Journals (Sweden)

A. Khajesarvi

2015-07-01

Full Text Available In the present study, nanocrystalline Ni50Al50-xMox (X = 0, 0.5, 1, 2.5, 5 intermetallic compound was produced through mechanical alloying of nickel, aluminum, and molybdenum powders. AlNi compounds with good and attractive properties such as high melting point, high strength to weight ratio and high corrosion resistance especially at high temperatures have attracted the attention of many researchers. Powders produced from milling were analyzed using scanning electron microscopy (SEM and X-ray diffractometry (XRD. The results showed that intermetallic compound of NiAl formed at different stage of milling operation. It was concluded that at first disordered solid solution of (Ni,Al was formed then it converted into ordered intermetallic compound of NiAl. With increasing the atomic percent of molybdenum, average grain size decreased from 3 to 0.5 μm. Parameter lattice and lattice strain increased with increasing the atomic percent of molybdenum, while the crystal structure became finer up to 10 nm. Also, maximum microhardness was obtained for NiAl49Mo1 alloy.

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-11-15

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

Recent prospects of MOX fuel and strategy about nuclear fuel cycle

International Nuclear Information System (INIS)

Liu Dingqin

1991-04-01

It is clearly described what is the preliminary adequate strategic concern for different nuclear power countries under different nuclear power development conditions. It is also stressed on the basic situation of the design technology, manufacture technology, operation experiences and quantitative economic analysis for MOX fuel application since fast breed reactor commercialization has been delayed. The author specially proposed that in a short term China should adopt an intermediate storage strategy matched with the construction of a pilot reprocessing plant to prepare the technical basis for commercialized reprocessing plant later on and to follow the development of MOX fuel technology

Revised conceptual designs for the FMDP MOX fresh fuel transport package

International Nuclear Information System (INIS)

Ludwig, S.B.; Michelhaugh, R.D.; Shappert, L.B.; Chae, S.M.; Tang, J.S.

1998-03-01

The revised conceptual designs described in this document provide a foundation for the development and certification of final transport package designs that will be needed to support the disposition of surplus weapons-grade plutonium as mixed-oxide (MOX) fuel in commercial light-water reactors in the US. This document is intended to describe the revised package design concepts and summarize the results of preliminary analyses and assessments of two new concepts for fresh MOX fuel transport packages that have been developed by Oak Ridge National Laboratory during the past year in support of the Department of Energy/Office of Fissile Materials Disposition

Structure-based domain assignment in Leishmania infantum EndoG: characterization of a pH-dependent regulatory switch and a C-terminal extension that largely dictates DNA substrate preferences.

Science.gov (United States)

Oliva, Cristina; Sánchez-Murcia, Pedro A; Rico, Eva; Bravo, Ana; Menéndez, Margarita; Gago, Federico; Jiménez-Ruiz, Antonio

2017-09-06

Mitochondrial endonuclease G from Leishmania infantum (LiEndoG) participates in the degradation of double-stranded DNA (dsDNA) during parasite cell death and is catalytically inactive at a pH of 8.0 or above. The presence, in the primary sequence, of an acidic amino acid-rich insertion exclusive to trypanosomatids and its spatial position in a homology-built model of LiEndoG led us to postulate that this peptide stretch might act as a pH sensor for self-inhibition. We found that a LiEndoG variant lacking residues 145-180 is indeed far more active than its wild-type counterpart at pH values >7.0. In addition, we discovered that (i) LiEndoG exists as a homodimer, (ii) replacement of Ser211 in the active-site SRGH motif with the canonical aspartate from the DRGH motif of other nucleases leads to a catalytically deficient enzyme, (iii) the activity of the S211D variant can be restored upon the concomitant replacement of Ala247 with Arg and (iv) a C-terminal extension is responsible for the observed preferential cleavage of single-stranded DNA (ssDNA) and ssDNA-dsDNA junctions. Taken together, our results support the view that LiEndoG is a multidomain molecular machine whose nuclease activity can be subtly modulated or even abrogated through architectural changes brought about by environmental conditions and interaction with other binding partners. © The Author(s) 2017. Published by Oxford University Press on behalf of Nucleic Acids Research.

Continuous process of powder production for MOX fuel fabrication according to ''granat'' technology

International Nuclear Information System (INIS)

Morkovnikov, V.E.; Raginskiy, L.S.; Pavlinov, A.P.; Chernov, V.A.; Revyakin, V.V.; Varykhanov, V.S.; Revnov, V.N.

2000-01-01

During last years the problem of commercial MOX fuel fabrication for nuclear reactors in Russia was solved in a number of directions. The paper deals with the solution of the problem of creating a continuous pilot plant for the production of MOX fuel powders on the basis of the home technology 'Granat', that was tested before on a small-scale pilot-commercial batch-operated plant of the same name and confirmed good results. (authors)

Mol 7C/6

International Nuclear Information System (INIS)

Aberle, J.; Schleisiek, K.; Schmuck, I.; Schmidt, L.; Romer, O.; Weih, G.

1995-01-01

The Mol 7C/6 coolant blockage experiment in the Belgian BR2 reactor yielded results different from Mol 7C experiments with low burnup pins: At 10% burnup local failure is not self-limiting, but requires active systems for detection and scram. The Mol 7C series was finished in 1991. In each of the test bundles Mol 7C/4, /5 and /6, 30 Mk I pins pre-irradiated in KNK II were used. The central blockage consisted of enriched UO 2 covering 30 percent of the bundle cross-section, with a height of 40 mm. The most important system for timely detection of coolant blockages of the type studied in Mol 7C/6 is based on DND. (orig.)

Association between VEGF polymorphisms (936c/t, -460t/c and -634g/c) with haplotypes and coronary heart disease susceptibility.

Science.gov (United States)

Han, Xia; Liu, Lili; Niu, Jiamin; Yang, Jun; Zhang, Zengtang; Zhang, Zhiqiang

2015-01-01

Our aim was to investigate the association between single nucleotide polymorphisms (SNPs) of vascular endothelial growth factor (VEGF) and coronary heart disease (CHD) susceptibility in Chinese Han population. 144 CHD patients and 150 healthy individuals were enrolled in the study. Three SNPs (936C/T, -460T/C and -634G/C) of VEGF were chose and then were genotyped with Sequenom time-of-flight mass spectrometry (TOFMS). Odds ratio (OR) with 95% confidence interval (CI) were used to evaluate the association of genotypes and haplotypes and CHD susceptibility. The frequencies of -460T/C CC genotype (13.6%) was found higher in the case group than that of control group (6.7%), which indicated that CC genotype was a risk factor for CHD (OR=2.50, 95% CI=1.10-5.68). Correspondently, the C allele appeared to increase the risk of CHD (OR=1.54, 95% CI=1.07-2.22). For -634G/C polymorphism, the risk of the CC genotype carrier for CHD increased 2.24 fold compared to the wild genotype. Moreover, -634G/CC allele was significantly associated with CHD susceptibility (OR=1.65, 95% CI=1.15-2.36). In addition, +936C/T CT genotype and C allele appeared to be a genetic-susceptibility factors for CHD (OR=2.43, 95% CI=1.44-4.10; OR=1.95, 95% CI=1.26-3.02). The haplotype analysis showed that T-C-T, C-C-C and C-G-C haplotypes all could increase the risk for CHD (OR: 2.43, 2.77 and 2.33). we concluded VEGF polymorphisms were associated with CHD susceptibility. Moreover, the haplotypes of T-C-T, C-C-C and C-G-C all could increase the risk for CHD.

C-reactive protein + 1059 G>C polymorphism in type 2 diabetes and coronary artery disease patients.

Science.gov (United States)

Kaur, Ramandeep; Matharoo, Kawaljit; Sharma, Rubina; Bhanwer, A J S

2013-12-01

Human C-reactive protein (CRP) is an acute phase reactant involved in chronic and acute inflammation. CRP is associated with metabolic syndrome, obesity, atherosclerosis, unstable angina, insulin resistance and diabetes. The present study evaluates the association of + 1059 G>C silent polymorphism in exon 2 of CRP gene in 581 cases [CAD (206), T2D (266), T2D with CAD (109)] and 235 controls in the population of Punjab (North-West India). The frequency of + 1059 G allele is highest in CAD (98.3%) followed by T2D (98.1%), T2D + CAD cases (97.7%) and controls (94.7%). G-allele is associated with increased risk of T2D [P = 0.003, OR = 2.93 (1.39-6.17)] and CAD [P = 0.004, OR = 3.25 (1.39-7.60)] in comparison to controls. Recessive model shows that GG genotype increases the risk of CAD by 4 fold (P = 0.003, OR = 4.19, 1.62-10.80), T2D by 3 fold (P = 0.008, OR = 3.23, 1.36-7.60) and T2D + CAD by 3.5 fold (P = 0.029, OR = 3.64, 1.14-11.66). Factor analyses show that BMI, WC, and WHR are core predictors for CAD and T2D, whereas CHO, TG and VLDL for T2D + CAD. The present study concludes that GG genotype of CRP + 1059 G>C polymorphism and clustering of obesity and dyslipidemia underlie the risk towards CAD, T2D and T2D + CAD in the North-West Indian population of Punjab.

Effect of heat treatment on impact resistance of AU5GT and AS7G06 aluminum alloys

Energy Technology Data Exchange (ETDEWEB)

Muzamil, Muhammad; Akhtar, Maaz; Samiuddin, Muhammad; Mehdi, Murtuza [NED University of Engineering and Technology, Karachi (Pakistan)

2016-10-15

Impact strength is one of the major mechanical properties that a material should possess in order to absorb sudden changes in the load intensity. The objective of current study is to compare the impact strength of two material (AU5GT and AS7G06), which are used in different structural applications. Almost no work is available which compares the impact strength of selected grade alloys along with different heat treatment cycles. Specimens are heat treated first as per designed cycles, later impact testing is performed. Charpy impact test is conducted in accordance with ASTM E23-12 standard method on three samples with and without heat treatment for each cycle. Solutionizing on samples is done at constant time and temperature to achieve homogenization. Later, aging is conducted at different temperatures ranging from 100-200°C (different intervals) at constant time to find the effect of precipitation hardness that actually increases the strength. Sample hardness is determined using Vickers micro hardness testing machine for each heat treatment cycle. Charpy test results provided the impact energy that is used to determine the strength before fracture. Heat treated samples have showed increase in impact strength for AS7G06 aluminum alloy while AU5GT shows very little change. This is because of growing the precipitation with respect to temperature, which resulted in more hard regions across grains. Hardness also shows an increasing relationship, as expected. Fracture surfaces are analyzed on stereo microscopy and Scanning electron microscopy (SEM) to find the final mode of fracture, that is brittle, ductile or transitional (combination of both brittle and ductile)

Characterization of the human pH- and PKA-activated ClC-2G(2 alpha) Cl- channel.

Science.gov (United States)

Sherry, A M; Stroffekova, K; Knapp, L M; Kupert, E Y; Cuppoletti, J; Malinowska, D H

1997-08-01

A ClC-2G(2 alpha) Cl- channel was identified to be present in human lung and stomach, and a partial cDNA for this Cl- channel was cloned from a human fetal lung library. A full-length expressible human ClC-2G(2 alpha) cDNA was constructed by ligation of mutagenized expressible rabbit ClC-2G(2 alpha) cDNA with the human lung ClC-2G(2 alpha) cDNA, expressed in oocytes, and characterized at the single-channel level. Adenosine 3',5'-cyclic monophosphate-dependent protein kinase (PKA) treatment increased the probability of opening of the channel (Po). After PKA activation, the channel exhibited a linear (r = 0.99) current-voltage curve with a slope conductance of 22.1 +/- 0.8 pS in symmetric 800 mM tetraethylammonium chloride (TEACl; pH 7.4). Under fivefold gradient conditions of TEACl, a reversal potential of +21.5 +/- 2.8 mV was measured demonstrating anion-to-cation discrimination. As previously demonstrated for the rabbit ClC-2G(2 alpha) Cl- channel, the human analog, hClC-2G(2 alpha), was active at pH 7.4 as well as when the pH of the extracellular face of the channel (trans side of the bilayer; pHtrans) was asymmetrically reduced to pH 3.0. The extent of PKA activation was dependent on pHtrans. With PKA treatment, Po increased fourfold with a pHtrans of 7.4 and eightfold with a pHtrans of 3.0. Effects of sequential PKA addition followed by pHtrans reduction on the same channel suggested that the PKA- and pH-dependent increases in channel Po were separable and cumulative. Northern analysis showed ClC-2G(2 alpha) mRNA to be present in human adult and fetal lung and adult stomach, and quantitative reverse transcriptase-polymerase chain reaction showed this channel to be present in the adult human lung and stomach at about one-half the level found in fetal lung. The findings of the present study suggest that the ClC-2G(2 alpha) Cl- channel may play an important role in Cl- transport in the fetal and adult human lung.

Construction of stable Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} metal/non-metal nitride hybrids with enhanced visible-light photocatalysis

Energy Technology Data Exchange (ETDEWEB)

Jiang, Yinhua, E-mail: yms418@126.com [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 2120013,PR China (China); Liu, Peipei; Chen, YeCheng; Zhou, Zhengzhong; Yang, Haijian [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 2120013,PR China (China); Hong, Yuanzhi; Li, Fan; Ni, Liang [School of Materials Science and Engineering, Jiangsu University, Zhenjiang 2120013,PR China (China); Yan, Yongsheng [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 2120013,PR China (China); Gregory, Duncan H, E-mail: duncan.gregory@glasgow.ac.uk [School of Chemistry, University of Glasgow, Glasgow G12 8QQ (United Kingdom)

2017-01-01

Highlights: • Novel Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} metal/non-metal nitride hybrids were synthesized. • The hybrid nitrides showed enhanced visible-light photocatalytic performance. • The Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} hybrid nitride exhibited excellent photostability. • The hole is the main photoactive specie for the degradation of RhB. - Abstract: In this paper, a novel Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} metal/non-metal nitride hybrid was successfully synthesized by a facile impregnation method. The photocatalytic activity of Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} hybrid nitrides was evaluated by the degradation of organic dye rhodamine B (RhB) under visible light irradiation, and the result indicated that all Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} samples exhibited distinctly enhanced photocatalytic activities for the degradation of RhB than pure g-C{sub 3}N{sub 4}. The optimal Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} composite sample, with Ta{sub 3}N{sub 5} mass ratio of 2%, demonstrated the highest photocatalytic activity, and its degradation rate constant was 2.71 times as high as that of pure g-C{sub 3}N{sub 4}. The enhanced photocatalytic activity of this Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} metal/metal-free nitride was predominantly attributed to the synergistic effect which increased visible-light absorption and facilitated the efficient separation of photoinduced electrons and holes. The Ta{sub 3}N{sub 5}/g-C{sub 3}N{sub 4} hybrid nitride exhibited excellent photostability and reusability. The possible mechanism for improved photocatalytic performance was proposed. Overall, this work may provide a facile way to synthesize the highly efficient metal/metal-free hybrid nitride photocatalysts with promising applications in environmental purification and energy conversion.

Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

International Nuclear Information System (INIS)

Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

1998-01-01

This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

Nonuniform transformation field analysis of multiphase elasto viscoplastic materials: application to MOX fuels

International Nuclear Information System (INIS)

Roussette, S.

2005-05-01

The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elasto viscoplastic materials. Karhunen-Loeve expansion is proposed to optimize the plastic modes. Then the method is extended to porous elasto viscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D. (author)

Plasminogen activator inhibitor-1 4G/5G polymorphism in infertile women with and without endometriosis.

Science.gov (United States)

Gonçalves-Filho, Rubens P; Brandes, Ariel; Christofolini, Denise M; Lerner, Tatiana G; Bianco, Bianca; Barbosa, Caio P

2011-05-01

To evaluate PAI-1 genotypes in a group of infertile women with or without endometriosis and control subjects. Case-control study. Human Reproduction Center of Medicina do ABC Faculty. One hundred and forty infertile women with endometriosis, 64 women with idiopathic infertility and 148 fertile women as control subjects. The PAI-1 4G/5G polymorphism was identified by restriction fragment length polymorphism-polymerase chain reaction. Genotype distribution and allele frequency of the 4G/5G polymorphism of the PAI-1 gene. The frequencies of genotypes 4G/4G, 4G/5G and 5G/5G of the PAI-1 gene in the infertile women with endometriosis were 38.6, 37.1 and 24.3%, respectively, and in the control group 24.3, 33.8 and 41.9%, respectively (p=0.003). When the infertile women with endometriosis were divided according to their endometriosis stage, genotypes 4G/4G, 4G/5G and 5G/5G were identified, respectively, in 36.7, 32.9 and 30.4% of the patients with minimal/mild endometriosis (p=0.102) and in 41.0, 42.6 and 16.4% of the patients with moderate/severe endometriosis (p=0.001); in the women with idiopathic infertility, these genotypes were found at a frequency of 29.7, 34.3 and 36%, respectively (p=0.637). The data suggest that, in Brazilian women, the PAI-1 4G/5G polymorphism may be associated with a risk of endometriosis-associated infertility. © 2011 The Authors Acta Obstetricia et Gynecologica Scandinavica© 2011 Nordic Federation of Societies of Obstetrics and Gynecology.

Study on transport safety of refresh MOX fuel. Radiation dose from package hypothetically submerged into sea

International Nuclear Information System (INIS)

Tsumune, Daisuke; Suzuki; Hiroshi; Saegusa, Toshiari; Maruyama, Koki; Ito, Chihiro; Watabe, Naoto

1999-01-01

The sea transport of fresh MOX fuel from Europe to Japan is under planning. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel will be transported safely on the sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unexpected reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

FBXO7 Y52C polymorphism as a potential protective factor in Parkinson's disease.

Directory of Open Access Journals (Sweden)

Chiung-Mei Chen

Full Text Available Mutations in the F-box only protein 7 gene (FBXO7, the substrate-specifying subunit of SCF E3 ubiquitin ligase complex, cause Parkinson's disease (PD-15 (PARK15. To identify new variants, we sequenced FBXO7 cDNA in 80 Taiwanese early onset PD patients (age at onset ≤ 50 and only two known variants, Y52C (c.155A>G and M115I (c.345G>A, were found. To assess the association of Y52C and M115I with the risk of PD, we conducted a case-control study in a cohort of PD and ethnically matched controls. There was a nominal difference in the Y52C G allele frequency between PD and controls (p = 0.045. After combining data from China [1], significant difference in the Y52C G allele frequency between PD and controls (p = 0.012 and significant association of G allele with decreased PD risk (p = 0.017 can be demonstrated. Upon expressing EGFP-tagged Cys52 FBXO7 in cells, a significantly reduced rate of FBXO7 protein decay was observed when compared with cells expressing Tyr52 FBXO7. In silico modeling of Cys52 exhibited a more stable feature than Tyr52. In cells expressing Cys52 FBXO7, the level of TNF receptor-associated factor 2 (TRAF2 was significantly reduced. Moreover, Cys52 FBXO7 showed stronger interaction with TRAF2 and promoted TRAF2 ubiquitination, which may be responsible for the reduced TRAF2 expression in Cys52 cells. After induced differentiation, SH-SY5Y cells expressing Cys52 FBXO7 displayed increased neuronal outgrowth. We therefore hypothesize that Cys52 variant of FBXO7 may contribute to reduced PD susceptibility in Chinese.

Overview of safeguards aspects related to MOX fuel

International Nuclear Information System (INIS)

Heinonen, O.J.; Murakami, K.; Shea, T.

2000-01-01

Recent developments in the light of the IAEA verification requirements for MOX fuel at reactors and bulk handling facilities are discussed. Impact of the Additional Protocol and Integrated Safeguards System is briefly addressed. Agency's work undertaken with regard to the nuclear arms control and reduction is presented. (author)

Fresh MOX fuel transport in Germany: experience for using the MX6

Energy Technology Data Exchange (ETDEWEB)

Lallemant, T. [COGEMA Logistics (AREVA Group), Bagnols/sur Ceze (France); Marien, L. [FBFC-I (AREVA Group), Dessel (Belgium); Wagner, R. [RWE, Gundremmingen (Germany); Jahreiss, W. [FRAMATOME ANP GmbH (AREVA Group), Erlangen (Germany); Tschiesche, H. [NCS, Hanau (Germany)

2004-07-01

The MX6 packaging developed by COGEMA LOGISTICS replaces the BWR SIEMENS packaging and SIEMENS III packaging for the transport of either BWR or PWR fresh MOX assemblies. It is licensed in France, Germany and Belgium according to TS-R-1 requirements (IAEA 1996). The associated security transport system was developed in co-operation with NCS (Nuclear Cargo + Service GmbH). The MX6 packaging is based on innovative solutions implemented at each step of the design. In 2004, RWE GUNDREMMINGEN Nuclear Power Plant (NPP) will be the first NPP delivered with the MX6 system and MOX assemblies manufactured by BELGONUCLEAIRE and FBFC in Belgium. Before this first transport, successful cold tests were performed for qualification of the whole system with the participation of all parties involved: NPP, carrier, fuel supplier and local Authorities. These tests were conducted by the NPP's operators in FBFC and GUNDREMMINGEN facilities and lead to the validation of the operating manual. Specific conditions for the return of the empty MX6 were also agreed between all parties. Similar operation will be conducted in each NPP before the first use of the MX 6. The large payload of the MX6: - 16 BWR MOX assemblies in one packaging instead of 2 - 6 PWR MOX assemblies in one packaging instead of 3 contributes to the optimisation of the dose uptake during unloading in the NPP. In this paper, the main contributors to the first MOX transport to Germany with the MX6 will present their involvement and feedback at each step of the transport of this new type of packaging, including loading and unloading operations. The use of the MX6 will be extended to other German NPP's from the next year. After FBFC in Belgium, MELOX in France will load the MX6 as well as the current MX8 packaging for the delivery to the French NPP's.

Fresh MOX fuel transport in Germany: experience for using the MX6

International Nuclear Information System (INIS)

Lallemant, T.; Marien, L.; Wagner, R.; Jahreiss, W.; Tschiesche, H.

2004-01-01

The MX6 packaging developed by COGEMA LOGISTICS replaces the BWR SIEMENS packaging and SIEMENS III packaging for the transport of either BWR or PWR fresh MOX assemblies. It is licensed in France, Germany and Belgium according to TS-R-1 requirements (IAEA 1996). The associated security transport system was developed in co-operation with NCS (Nuclear Cargo + Service GmbH). The MX6 packaging is based on innovative solutions implemented at each step of the design. In 2004, RWE GUNDREMMINGEN Nuclear Power Plant (NPP) will be the first NPP delivered with the MX6 system and MOX assemblies manufactured by BELGONUCLEAIRE and FBFC in Belgium. Before this first transport, successful cold tests were performed for qualification of the whole system with the participation of all parties involved: NPP, carrier, fuel supplier and local Authorities. These tests were conducted by the NPP's operators in FBFC and GUNDREMMINGEN facilities and lead to the validation of the operating manual. Specific conditions for the return of the empty MX6 were also agreed between all parties. Similar operation will be conducted in each NPP before the first use of the MX 6. The large payload of the MX6: - 16 BWR MOX assemblies in one packaging instead of 2 - 6 PWR MOX assemblies in one packaging instead of 3 contributes to the optimisation of the dose uptake during unloading in the NPP. In this paper, the main contributors to the first MOX transport to Germany with the MX6 will present their involvement and feedback at each step of the transport of this new type of packaging, including loading and unloading operations. The use of the MX6 will be extended to other German NPP's from the next year. After FBFC in Belgium, MELOX in France will load the MX6 as well as the current MX8 packaging for the delivery to the French NPP's

5G CHAMPION - rolling out 5G in 2018

OpenAIRE

Mueck, M.; Strinati, E.C.; Kim, I.-G.; Clemente, A.; Dore, J.-B.; Domenico, A. de; Kim, T.; Choi, T.; Chung, H.K.; Destino, G.; Pärssinen, A.; Pouttu, A.; Latva-aho, M.; Chuberre, N.; Gineste, M.

2016-01-01

The 5G CHAMPION Consortium will provide the first fully integrated and operational 5G prototype in 2018 - this effort is a major leap ahead compared to existing punctual technology trials, such as, e.g., Proof-of-Concept platforms focusing on mm Wave communication in specific bands, etc. This paper describes the overall set-up including a synergetic combination of technologies such as beamforming based mm Wave & Satellite service provisioning, virtualized infrastructure, software reconfigurat...

Synthesis, DFT band structure calculations, optical and photoelectrical characterizations of the novel 5-hydroxy-4-methoxy-7-oxo-7H-furo[3,2-g]chromene-6-carbonitrile (HMOFCC)

Science.gov (United States)

Ibrahim, Magdy A.; Halim, Shimaa Abdel; Roushdy, N.; Farag, A. A. M.; El-Gohary, Nasser M.

2017-11-01

Reaction of 4-methoxy-5-oxo-5H-furo[3,2-g]chromene-6-carboxaldehyde (1) with hydroxylamine hydrochloride resulted in ring transformation producing the novel 5-hydroxy-4-methoxy-7-oxo-7H-furo[3,2-g]chromene-6-carbonitrile (HMOFCC). The structure was deduced based on its correct elemental analysis and spectral data (IR, 1H NMR, 13C NMR and mass spectra). The geometries of the HMOFCC were completely optimized by means of DFT-B3LYP/6-311++G (d,p) theoretical level. The ground state properties such as; total energy, the energy of HOMO and LUMO and Mulliken atomic charges were also determined. In addition, the two solvents; polar (methanol) and nonpolar (dioxane) were utilized to extract the electronic absorption spectra. The assignment of the detected bands was discussed by TD-DFT calculations. A cauliflower-like, as well as, needle-like leaves morphologies were observed using scanning electron microscope images. Two direct optical band gaps were extracted from the photon energy dependence of absorption coefficient at the band edges and found to be 1.16 and 2.56 eV. A characteristic emission peak of photoluminescence spectrum was observed and shifted depending on the solvent type. A remarkable rectification characteristic of HMOFCC/p-Si heterojunction confirms the diode-like behavior. The main important parameters like series resistance, shunt resistance and reverse saturation current show illumination dependence under influence of the illumination intensity range 20-100 mW/cm2. The heterojunction based HMOFCC showed phototransient properties under various illumination intensities which give the recommendation for the studied heterojunction in the field of optoelectronic device application.

Differences in TCDD-elicited gene expression profiles in human HepG2, mouse Hepa1c1c7 and rat H4IIE hepatoma cells

Directory of Open Access Journals (Sweden)

Burgoon Lyle D

2011-04-01

Full Text Available Abstract Background 2,3,7,8-Tetrachlorodibenzo-p-dioxin (TCDD is an environmental contaminant that elicits a broad spectrum of toxic effects in a species-specific manner. Current risk assessment practices routinely extrapolate results from in vivo and in vitro rodent models to assess human risk. In order to further investigate the species-specific responses elicited by TCDD, temporal gene expression responses in human HepG2, mouse Hepa1c1c7 and rat H4IIE cells were compared. Results Microarray analysis identified a core set of conserved gene expression responses across species consistent with the role of AhR in mediating adaptive metabolic responses. However, significant species-specific as well as species-divergent responses were identified. Computational analysis of the regulatory regions of species-specific and -divergent responses suggests that dioxin response elements (DREs are involved. These results are consistent with in vivo rat vs. mouse species-specific differential gene expression, and more comprehensive comparative DRE searches. Conclusions Comparative analysis of human HepG2, mouse Hepa1c1c7 and rat H4IIE TCDD-elicited gene expression responses is consistent with in vivo rat-mouse comparative gene expression studies, and more comprehensive comparative DRE searches, suggesting that AhR-mediated gene expression is species-specific.

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

International Nuclear Information System (INIS)

O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

1998-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-08-01

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

Triple helical polynucleotidic structures: an FTIR study of the C+ .G. Ctriplet.

Science.gov (United States)

Akhebat, A; Dagneaux, C; Liquier, J; Taillandier, E

1992-12-01

Triple helixes containing one homopurine poly dG or poly rG strand and two homopyrimidine poly dC or poly rC strands have been prepared and studied by FTIR spectroscopy in H2O and D2O solutions. The spectra are discussed by comparison with those of the corresponding third strands (auto associated or not) and of double stranded poly dG.poly dC and poly rG.poly rC in the same concentration range and salt conditions. The triplex formation is characterized by the study of the base-base interactions reflected by changes in the spectral domain involving the in-plane double bond vibrations of the bases. Modifications of the initial duplex conformation (A family form for poly rG.poly rC, B family form for poly dG.poly dC) when the triplex is formed have been investigated. Two spectral domains (950-800 and 1450-1350 cm-1) containing absorption bands markers of the N and S type sugar geometries have been extensively studied. The spectra of the triplexes prepared starting with a double helix containing only riboses (poly rC+.poly rG.poly rC and poly dC+.poly rG.poly rC) as well as that of poly rC+.poly dG.poly dC present exclusively markers of the North type geometry of the sugars. On the contrary in the case of the poly dC+.poly dG.poly dC triplex both N and S type sugars are shown to coexist. The FTIR spectra allow us to propose that in this case the sugars of the purine (poly dG) strand adopt the S type geometry.

FERMI-LAT OBSERVATIONS OF SUPERNOVA REMNANT G5.7–0.1, BELIEVED TO BE INTERACTING WITH MOLECULAR CLOUDS

Energy Technology Data Exchange (ETDEWEB)

Joubert, Timothy; Slane, Patrick [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Castro, Daniel [MIT-Kavli Center for Astrophysics and Space Research, 77 Massachusetts Avenue, Cambridge, MA, 02139 (United States); Gelfand, Joseph [NYU Abu Dhabi, P.O. Box 129188, Abu Dhabi (United Arab Emirates)

2016-01-10

This work reports on the detection of γ-ray emission coincident with the supernova remnant (SNR) G5.7–0.1 using data collected by the Large Area Telescope on board the Fermi Gamma-ray Space Telescope. The SNR is believed to be interacting with molecular clouds, based on 1720 MHz hydroxyl (OH) maser emission observations in its direction. This interaction is expected to provide targets for the production of γ-ray emission from π{sup 0}-decay. A γ-ray source was observed in the direction of SNR G5.7–0.1, positioned near the bright γ-ray source SNR W28. We model the emission from radio to γ-ray energies using a one-zone model. Following consideration of both π{sup 0}-decay and leptonically dominated emission scenarios for the MeV–TeV source, we conclude that a considerable component of the γ-ray emission must originate from the π{sup 0}-decay channel. Finally, constraints were placed on the reported ambiguity of the SNR distance through X-ray column density measurements made using XMM-Newton observations. We conclude G5.7–0.1 is a significant γ-ray source positioned at a distance of ∼3 kpc with luminosity in the 0.1–100 GeV range of L{sub γ} ≈ 7.4 × 10{sup 34} erg s{sup −1}.

Development and validation of the ENIGMA code for MOX fuel performance modelling

International Nuclear Information System (INIS)

Palmer, I.; Rossiter, G.; White, R.J.

2000-01-01

The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

2007-07-01

Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

Performance of cladding on MOX fuel with low 240Pu/239Pu ratio

International Nuclear Information System (INIS)

McCoy, K.; Blanpain, P.; Morris, R.

2015-01-01

The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world's first commercial irradiation of MOX fuel with a 240 Pu/ 239 Pu ratio less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This paper discusses the results of those examinations with emphasis on cladding performance. Exams relevant to the cladding included visual and eddy current exams, profilometry, microscopy, hydrogen analysis, gallium analysis, and mechanical testing. There was no discernible effect of the type of MOX fuel on the performance of the cladding. (authors)

5G in Open-Pit Mines

DEFF Research Database (Denmark)

Portela Lopes de Almeida, Erika; Caldwell, George; Rodriguez Larrad, Ignacio

2017-01-01

5G will play a pivotal role in the digitization of the industrial sector and is expected to make the best use of every bit of spectrum available. In this light, this paper presents the results of an extensive measurement campaign in two iron-ore open-pit mining complexes, at the 700 MHz and 2.6 GHz...... for the future wireless network design, simulation and performance evaluation. The results show that, in order to comply with ultra-reliable communications (URC) availability requirements, larger shadowing margins will have to be considered in the network planning in open-pit mines, when compared to traditional...

12 CFR 563g.7 - Form, content, and accounting.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Form, content, and accounting. 563g.7 Section... § 563g.7 Form, content, and accounting. (a) Form and content. Any offering circular or amendment filed... which they are made, not misleading. (b) Accounting requirements. To be declared effective an offering...

Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

International Nuclear Information System (INIS)

Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

2004-01-01

The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

Influence of plutonium contents in MOX fuel on destructive forces at fuel failure in the NSRR experiment

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

In order to confirm safety margins of the Mixed Oxide (MOX) fuel use in LWRs, pulse irradiation tests are planned in the Nuclear Safety Research Reactor (NSRR) with the MOX fuel with plutonium content up to 12.8%. Impacts of the higher plutonium contents on safety of the reactivity-initiated-accident (RIA) tests are examined in terms of generation of destructive forces to threat the integrity of test capsules. Pressure pulses would be generated at fuel rod failure by releases of high pressure gases. The strength of the pressure pulses, therefore, depends on rod internal - external pressure difference, which is independent to plutonium content of the fuel. The other destructive forces, water hammer, would be generated by thermal interaction between fuel fragments and coolant water. Heat flux from the fragments to the water was calculated taking account of changes in thermal properties of MOX fuels at higher plutonium contents. The results showed that the heat transfer from the MOX fuel would be slightly smaller than that from UO{sub 2} fuel fragments at similar size in a short period to cause the water hammer. Therefore, the destructive forces were not expected to increase in the new tests with higher plutonium content MOX fuels. (author)

On the Secrecy Capacity of 5G New Radio Networks

Directory of Open Access Journals (Sweden)

Ke Xiao

2018-01-01

Full Text Available The new radio technology for the fifth-generation wireless system has been extensively studied all over the world. Specifically, the air interface protocols for 5G radio access network will be standardized by the 3GPP in the coming years. In the next-generation 5G new radio (NR networks, millimeter wave (mmWave communications will definitely play a critical role, as new NR air interface (AI is up to 100 GHz just like mmWave. The rapid growth of mmWave systems poses a variety of challenges in physical layer (PHY security. This paper investigates those challenges in the context of several 5G new radio communication technologies, including multiple-input multiple-output (MIMO and nonorthogonal multiple access (NOMA. In particular, we introduce a ray-tracing (RT based 5G NR network channel model and reveal that the secrecy capacity in mmWave band widely depends on the richness of radio frequency (RF environment through numerical experiments.

Millimetre-Wave Backhaul for 5G Networks: Challenges and Solutions

Directory of Open Access Journals (Sweden)

Wei Feng

2016-06-01

Full Text Available The trend for dense deployment in future 5G mobile communication networks makes current wired backhaul infeasible owing to the high cost. Millimetre-wave (mm-wave communication, a promising technique with the capability of providing a multi-gigabit transmission rate, offers a flexible and cost-effective candidate for 5G backhauling. By exploiting highly directional antennas, it becomes practical to cope with explosive traffic demands and to deal with interference problems. Several advancements in physical layer technology, such as hybrid beamforming and full duplexing, bring new challenges and opportunities for mm-wave backhaul. This article introduces a design framework for 5G mm-wave backhaul, including routing, spatial reuse scheduling and physical layer techniques. The associated optimization model, open problems and potential solutions are discussed to fully exploit the throughput gain of the backhaul network. Extensive simulations are conducted to verify the potential benefits of the proposed method for the 5G mm-wave backhaul design.

MOX Lead Assembly Fabrication at the Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

Geddes, R.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Spiker, D.L.; Poon, A.P.

1997-12-01

The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

Early development of spinodal decomposition in neutron-irradiated Fe-35.5Ni-7.5Cr at 5500C

International Nuclear Information System (INIS)

Brager, H.R.; Garner, F.A.

1985-04-01

In Fe-35Ni-7.5Cr irradiated at 550 0 C to 2.5 x 10 22 n/cm 2 (E > 0.1 MeV) the spinodal decomposition observed at higher irradiation temperatures and higher neutron exposures is just beginning to form. The decomposition appears to begin very heterogeneously and may be assisted by the action of the inverse Kirkendall mechanism operating at various microstructural sinks

The MELOX MOX fabrication facility: history of an industrial success and future prospects

International Nuclear Information System (INIS)

Arslan, M.; Jacquet, R.; Krellmann, J.

2005-01-01

Along with the La Hague reprocessing plant, MELOX is part of the two industrial facilities that ensure the closure of the nuclear fuel cycle in France. Since started up in 1995, MELOX has specialized into recycling separated plutonium recovered from reprocessing operations performed at La Hague on spent UO 2 fuel. Capitalizing on the unique know-how acquired through thirty years of plutonium-based fuel fabrication at the Cadarache plant, this subsidiary of AREVA group has quickly become a worldwide expert in the industrial process of fabricating MOX: a fuel blend comprised of both uranium and plutonium oxides that allows at safely exploiting the energetic potential of plutonium. In order to address the various factors responsible for this industrial breakthrough, we will first present an overview of MELOX's history in regards of the emergence of a global MOX market. The added-value provided through treatment and recycling operations on spent fuel will be further described in terms of waste volume and radiotoxicity reduction. The emphasis will then be put on the total quality management policy that is at the core of MELOX's corporate strategy. Because MELOX has succeeded in meeting both productivity requirements and stringent quality constraints, it has won confidence from its European and Japanese clients. With increased production capacity of diversified MOX designs, MELOX is demonstrating the industrial efficiency of a new concept of MOX plants that is inspiring large construction projects in Japan, the US, and Russia. (authors)

Functional insight into the C-terminal extension of halolysin SptA from haloarchaeon Natrinema sp. J7.

Directory of Open Access Journals (Sweden)

Zhisheng Xu

Full Text Available Halolysin SptA from haloarchaeon Natrinema sp. J7 consists of a subtilisin-like catalytic domain and a C-terminal extension (CTE containing two cysteine residues. In this report, we have investigated the function of the CTE using recombinant enzymes expressed in Haloferax volcanii WFD11. Deletion of the CTE greatly reduced but did not abolish protease activity, which suggests that the CTE is not essential for enzyme folding. Mutational analysis suggests that residues Cys303 and Cys338 within the CTE form a disulfide bond that make this domain resistant to autocleavage and proteolysis under hypotonic conditions. Characterization of full-length and CTE-truncation enzymes indicates the CTE not only confers extra stability to the enzyme but also assists enzyme activity on protein substrates by facilitating binding at high salinities. Interestingly, homology modeling of the CTE yields a β-jelly roll-like structure similar to those seen in Claudin-binding domain of Clostridium perfringens enterotoxin (clostridial C-CPE and collagen binding domain (CBD, and the CTE also possesses collagen-binding activity, making it a potential candidate as an anchoring unit in drug delivery systems.

The presence of PAI-1 4G/5G and ACE DD genotypes increases the risk of early-stage AVF thrombosis in hemodialysis patients.

Science.gov (United States)

Güngör, Yahya; Kayataş, Mansur; Yıldız, Gürsel; Özdemir, Öztürk; Candan, Ferhan

2011-01-01

In this study, we investigated the relationship between early arteriovenous fistula (AVF) thrombosis with angiotensin-converting enzyme (ACE) gene and thrombophilic factor gene polymorphisms. Thirty-five patients who suffered from three or more fistula thrombosis episodes in the early period after AVF operation and 33 control patients with no history of thrombosis for at least 3 years were enrolled in this study. Factor V G1691A Leiden, factor V H1299R (R2), prothrombin G20210A, factor XIIIV34L, β-fibrinogen-455 G-A, glycoprotein IIIa L33P human platelet antigens (HPA-1), methylenetetrahydrofolate reductase C677T, and methylenetetrahydrofolate reductase A1298C gene polymorphisms were similar in both groups (p > 0.05). Plasminogen activator inhibitor 1 (PAI-1) 4G/5G genotype in the study group and 4G/4G genotype in the control group were significantly higher (p = 0.014). No significant difference was detected in terms of the 5G/5G genotype. With regard to the ACE gene polymorphism, the control group showed more ID genotype (19/33, 57.6%), whereas the study group showed more DD genotype (17/35, 48.6%). II genotype was similar in both groups (x(2) = 7.40, p = 0.025). The rate of ACE inhibitor-angiotensin II receptor blockers use was 5/35 in the study group (14.3%) and 5/33 in the control group (15.2%). Individuals with PAI-1 4G/5G genotype showed 5.03 times more risk of thrombosis when compared with 4G/4G and 5G/5G genotypes [p = 0.008, OR = 5.03, 95% confidence interval (1.44:17.64)]. Individuals with ACE DD genotype showed 4.25 times more risk of thrombosis when compared with II and ID [p = 0.008, OR = 4.25, 95% confidence interval (1.404:12.83)]. PAI-1 4G/5G and ACE DD genotypes are associated with increased risk for early AVF thrombosis.

The use of commercial microwave dissolution equipment for the fast and reliable dissolution of high-fired POX and MOX samples

International Nuclear Information System (INIS)

Tushingham, J.; McInnes, C.; Firkin, S.

1998-09-01

The use of commercially available microwave dissolution equipment for the fast and reliable dissolution of high-fired plutonium dioxide (POX) and mixed oxide (MOX) samples has been evaluated for application to Safeguards Analysis. Under the auspices of the UK R and D Support Programme to the IAEA, equipment has been purchased and tested for the high-pressure microwave dissolution of POX samples fired to 1250 deg. C and MOX samples fired to 1600 deg. C, in concentrated nitric acid and hydrofluoric acid mixture. Considerable problems were encountered during development of procedures for microwave dissolution, resulting largely from sudden changes in pressure within dissolution vessels, which resulted in actuation of safety interlocks designed to prevent overpressurisation. These difficulties were alleviated by controlling the microwave power to reduce the reaction temperature and pressure, and also by introducing additional safety valves into the digestion vessels. Using microwave digestion, dissolution times for high fired POX and MOX samples were substantially reduced. Samples which required ca. 10 hours to dissolve by conventional means could be dissolved in ca. 80 minutes by microwave digestion. Whilst a similar performance in terms of plutonium recovery was achieved for some materials by microwave and conventional dissolution, for other materials microwave dissolution gave higher plutonium recoveries but with poorer precision. This suggests the possible presence of some plutonium oxide within high-fired materials which is more difficult to dissolve than the bulk, and which is perhaps dissolved to an additional but variable degree by the current microwave dissolution procedure. Microwave dissolution has been demonstrated to increase the speed of dissolution of high-fired POX and MOX materials, compared with conventional dissolution. However, the technique has not yet proved satisfactory for the complete dissolution of all high-fired materials tested because of

Dose assessment for public at the hypothetical submergence of a fresh MOX fuel package

International Nuclear Information System (INIS)

Tsumune, Daisuke; Saegusa, Toshiari; Suzuki, Hiroshi; Maruyama, Koki

2000-01-01

For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel is transported safety on the sea. To gain the public acceptance for this transport, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unknown reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

Use of the program TNHXY in assemblies type MOX in comparison with CASMO-4; Utilizacion del programa TNHXY en ensambles tipo MOX en comparacion con CASMO-4

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Enriquez C, P. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

2011-11-15

In this work a comparison is made in the analysis of fuel assemblies type MOX among the CASMO-4 code and the program TNHXY (Transport of neutrons with Hybrid Nodal schemes in X Y geometry) which solves the equation of neutrons transport in stationary state and X Y geometry using nodal schemes type finite element -hybrid-, such named in correspondence to the parameters that interpolate. The program TNHXY has been validated previously by means of different test problems or benchmark that some authors have solved using other numeric techniques. In addition to analyzing assemblies type BWR. Some of the codes with which have been realized the validations are TWOTRAN as well as other commercial codes as, Helios, MCNP-4B and Cpm-3. The reason of to do this comparative is to able to observe the versatility of the program TNHXY with regard to CASMO-4 relating to the assemblies analysis type MOX and BWR, offering an alternative in the analysis of the same assemblies and with this comparison is confirmed even more the program TNHXY. For the comparison was analyzed a fuel assembly of the type GNF2 for a reactor type BWR that contains MOX with 10 enrichment types for a specific burnt pass. (Author)

The high moderating ratio reactor using 100% MOX reloads

International Nuclear Information System (INIS)

Barbrault, P.

1994-06-01

This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

Design and Discovery of Functionally Selective Serotonin 2C (5-HT2C) Receptor Agonists.

Science.gov (United States)

Cheng, Jianjun; McCorvy, John D; Giguere, Patrick M; Zhu, Hu; Kenakin, Terry; Roth, Bryan L; Kozikowski, Alan P

2016-11-10

On the basis of the structural similarity of our previous 5-HT 2C agonists with the melatonin receptor agonist tasimelteon and the putative biological cross-talk between serotonergic and melatonergic systems, a series of new (2,3-dihydro)benzofuran-based compounds were designed and synthesized. The compounds were evaluated for their selectivity toward 5-HT 2A , 5-HT 2B , and 5-HT 2C receptors in the calcium flux assay with the ultimate goal to generate selective 5-HT 2C agonists. Selected compounds were studied for their functional selectivity by comparing their transduction efficiency at the G protein signaling pathway versus β-arrestin recruitment. The most functionally selective compound (+)-7e produced weak β-arrestin recruitment and also demonstrated less receptor desensitization compared to serotonin in both calcium flux and phosphoinositide (PI) hydrolysis assays. We report for the first time that selective 5-HT 2C agonists possessing weak β-arrestin recruitment can produce distinct receptor desensitization properties.

Neonatal Hyperbilirubinemia in infants with G6PD c.563C > TVariant

Directory of Open Access Journals (Sweden)

Moiz Bushra

2012-08-01

Full Text Available Abstract Background There is a strong correlation between glucose-6-phosphate dehydrogenase (G6PD deficiency and neonatal hyperbilirubinemia with a rare but potential threat of devastating acute bilirubin encephalopathy. G6PD deficiency was observed in 4–14% of hospitalized icteric neonates in Pakistan. G6PD c.563C > T is the most frequently reported variant in this population. The present study was aimed at evaluating the time to onset of hyperbilirubinemia and the postnatal bilirubin trajectory in infants having G6PD c.563C > T. Methods This was a case–control study conducted at The Aga Khan University, Pakistan during the year 2008. We studied 216 icteric male neonates who were re-admitted for phototherapy during the study period. No selection was exercised. Medical records showed that 32 were G6PD deficient while 184 were G6PD normal. Each infant was studied for birth weight, gestational age, age at the time of presentation, presence of cephalhematoma, sepsis and neurological signs, peak bilirubin level, age at peak bilirubin level, days of hospitalization, whether phototherapy or exchange blood transfusion was initiated, and the outcome. During hospital stay, each baby was tested for complete blood count, reticulocyte count, ABO and Rh blood type, direct antiglobulin test and quantitative G6PD estimation [by kinetic determination of G6PDH]. G6PDgenotype was analyzed in 32 deficient infants through PCR-RFLP analysis and gene sequencing. Results G6PD variants c.563C > T and c.131 C > G were observed in 21 (65% and three (9% of the 32 G6PD deficient infants, respectively. DNA of eight (25% newborns remained uncharacterized. In contrast to G6PD normal neonates, infants with c.563C > T variant had significantly lower enzyme activity (mean ± 1SD; 0.3 ± 0.2 U/gHb vs. 14.0 ± 4.5 U/gHb, p p = 0.008 which peaked earlier after birth (mean ± 1SD 2.9 ± 1.6 vs. 4.3 ± 2.3 days, p = 0.007. No statistically significant

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

International Nuclear Information System (INIS)

Johnson, E.R.; Best, R.E.

2009-01-01

The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the 'UREX+3c fuel cycle' and the 'Alternative Fuel Cycle' (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the

Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

Energy Technology Data Exchange (ETDEWEB)

E. R. Johnson; R. E. Best

2009-12-28

The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

Cross sections for 12C+12C→12C(0+2)+12C(g.s.) using breathing mode doorways

International Nuclear Information System (INIS)

Ahmed, M.U.; Beres, W.P.

1982-01-01

A previously derived projection operator method is applied to the calculation of the cross section for 12 C+ 12 C→ 12 C(0 + 2 )+ 12 C(g.s.) with a breathing mode model being used to describe the 0 + 2 (7.68 MeV) state of 12 C. The relationship to processes leading to alpha particle channels is discussed. The cross section for 12 C+ 12 C→ 12 C(3 - )+ 12 C(g.s.) is also calculated and possible correlations with inelastic scattering to the 0 + 2 and 2 + states of 12 C are discussed. The results for both 0 + 2 and 3 - inelastic scattering are in reasonable agreement with experiment

The C(-1019G 5-HT1A promoter polymorphism and personality traits: no evidence for significant association in alcoholic patients

Directory of Open Access Journals (Sweden)

Zill P

2006-02-01

Full Text Available Abstract The 5HT1A receptor is one of at least 14 different receptors for serotonin which has a role in moderating several brain functions and may be involved in the aetiology of several psychiatric disorders. The C(-1019G 5-HT1A promoter polymorphism was reported to be associated with major depression, depression-related personality traits and suicidal behavior in various samples. The G(-1019 allele carriers are prone to depressive personality traits and suicidal behavior, because serotonergic neurotransmission is reduced. The aim of this study is to replicate previous findings in a sample of 185 Alcohol-dependent individuals. Personality traits were evaluated using the NEO FFI and TCI. History of suicidal behavior was assessed by a standardized semistructured interview (SSAGA. No significant differences across C(-1019G 5-HT1A genotype groups were found for TCI temperament and character traits and for NEO FFI personality scales. No association was detected between this genetic variant and history of suicide attempts. These results neither support a role of C(-1019G 5-HT1A promoter polymorphism in the disposition of personality traits like harm avoidance or neuroticism, nor confirm previous research reporting an involvement of the G allele in suicidal behavior in alcoholics. Significant associations, however, were detected between Babor's Type B with number of suicide attempts in history, high neuroticism and harm avoidance scores in alcoholics.

Effect of the hydration temperature on the microstructure of Class G cement: C-S-H composition and density

International Nuclear Information System (INIS)

Bahafid, Sara; Ghabezloo, Siavash; Duc, Myriam; Faure, Pamela; Sulem, Jean

2017-01-01

Curing temperature has a significant influence on cement paste microstructure and the properties of its principal hydrate C-S-H. In this paper, the effect of the hydration temperature in the range of 7 °C to 90 °C on the microstructure of a class G oil-well cement is studied. This is done by combining various experimental methods, including X-ray diffraction associated with the Rietveld analysis, thermo-gravimetric analysis, mercury intrusion porosimetry and porosity evaluation by drying. The experimental results show an increase of the capillary porosity and a decrease of the gel porosity by increasing the hydration temperature. This is attributed to a decrease of the C-S-H intrinsic porosity and a corresponding increase of the C-S-H density for higher curing temperatures. The experimental results are used in a simple analysis method to evaluate the density of C-S-H, as well as its C/S ratio and H/S ratio in dry and saturated conditions. The evaluated C-S-H density varies from 1.88 g/cm 3 at 7 °C to 2.10 g/cm 3 at 90 °C. The results also show a decrease of molar C/S ratio with increasing hydration temperature from 1.93 at 7 °C to 1.71 at 90 °C and of the H/S ratio from 5.1 at 7 °C to 2.66 at 90 °C.

Synthesis and evaluation of [11C]Cimbi-806 as a potential PET ligand for 5-HT₇ receptor imaging

DEFF Research Database (Denmark)

Herth, Matthias Manfred; Hansen, Hanne Demant; Ettrup, Anders Janusz

2012-01-01

)-N,N-dimethylethanamine ([(11)C]Cimbi-806) as a radioligand for imaging brain 5-HT(7) receptors with positron emission tomography (PET). Precursor and reference compound was synthesized and subsequent (11)C-labelling with [(11)C]methyltriflate produced [(11)C]Cimbi-806 in specific activities ranging from 50 to 300 GBq...... of appropriate in vivo blocking with a 5-HT(7) receptor selective compounds renders the conclusion that [(11)C]Cimbi-806 is not an appropriate PET radioligand for imaging the 5-HT(7) receptor in vivo....

Chemical characterisation of MOX grinder sludge and process evaluation for its dry recycling

Energy Technology Data Exchange (ETDEWEB)

Mallik, G K; Fulzele, A K; Kothari, M; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility

1997-09-01

A large quantity of sludge (approximately 5%) is generated as a result of centreless grinding of MOX pellets. Plutonium and uranium are recovered from such sludge, consisting of coolant, resin and some metallic impurities, by a wet chemical route. A case has been made for the recycling of the sludge by an optimum dry route on the basis of chemical characterisation of sludge generated at Advanced Fuel Fabrication Facility using diamond grinding wheel. (author). 2 tabs.

Chemical characterisation of MOX grinder sludge and process evaluation for its dry recycling

International Nuclear Information System (INIS)

Mallik, G.K.; Fulzele, A.K.; Kothari, M.; Bhargava, V.K.; Kamath, H.S.

1997-01-01

A large quantity of sludge (approximately 5%) is generated as a result of centreless grinding of MOX pellets. Plutonium and uranium are recovered from such sludge, consisting of coolant, resin and some metallic impurities, by a wet chemical route. A case has been made for the recycling of the sludge by an optimum dry route on the basis of chemical characterisation of sludge generated at Advanced Fuel Fabrication Facility using diamond grinding wheel. (author). 2 tabs