bluuebunny/crossref_metadata_2025 huggingface dataset:
Clinical impact of an angiotensin I-converting enzyme insertion/deletion and kinin B2 receptor +9/-9 polymorphisms in the prognosis of renal transplantation
Authors: Amorim Carlos E.N. , Nogueira Eliana , Almeida Sandro S. , Gomes Pedro P.G. , Bacurau Reury F.P. , Ozaki K. Suzete , Cenedeze Marcos A. , Silva Filho Alvaro P. , Câmara Niels O.S. , Araujo Ronaldo C.
Published: February 2013

There is a consensus in the scientific literature that supports the importance of the kallikrein kinin and renin angiotensin systems in renal physiology, but few studies have investigated their importance after renal transplantation. The aim of this study was to investigate the clinical effects of the insertion/deletion polymorphism in the angiotensin I-converting enzyme (ACE) gene and the +9/-9 polymorphism in the kinin B2 receptor (B2R) gene in kidney-transplanted patients (n=215 ACE, n=203 B2R) compared with 443 healthy individuals. Demographic results showed that there is a higher frequency of the D allele (high plasma ACE activity) and +9 allele (lower B2R expression) in transplant patients compared with control individuals. We also observed a higher frequency of these alleles in patients who had an elevated level of plasma creatinine. At day 7 post-transplantation, we found a higher prevalence of individuals with the DD genotype with elevated plasma creatinine level. Furthermore, individuals with the DD genotype had a higher chronic allograft dysfunction and graft loss compared with the II patient genotype, which showed no loss of graft. Taken together, our data suggest that the DD genotype is an indicator of an unfavorable prognosis following renal transplantation and could be related to kinin modulation.
pKa
values and redox potentials of proteins. What do they mean?
Authors: Ullmann G. Matthias , Bombarda Elisa
Published: February 2013

In this article, we review a microstate model that uses protonation and redox microstates in order to understand the complex pH and redox titration of proteins and other polyelectrolytes. From this model, it becomes obvious that it is impossible to assign pK

a
values or redox potentials to individual protonatable or redox-active sites in a protein in which many of such sites interact. Instead each site is associated with many microscopic equilibrium constants that may lead to irregular or even non-monotonic titration curves of some groups. The microstate model provides a closed theoretical framework to discuss such phenomena.
Alternate pathways for NADH oxidation in Thermus thermophilus using type 2 NADH dehydrogenases
Authors: Venkatakrishnan Padmaja , Lencina Andrea M. , Schurig-Briccio Lici A. , Gennis Robert B.
Published: February 2013

Type 2 NADH dehydrogenase (NDH-2) is a single-subunit membrane-associated flavoenzyme that is part of the respiratory chain of many prokaryotes. The enzyme catalyzes the electron transfer from NADH to quinone but is not directly coupled to the generation of a proton motive force. The purpose of the current work is to compare two different NDH-2s that are encoded in strains of Thermus thermophilus. The aerobic T. thermophilus HB27 strain expresses one NDH-2 that has been previously isolated and characterized. In this work it is shown that a gene, which is misannotated as an NADH oxidase, encodes this enzyme. Unlike HB27, strain NAR1 of T. thermophilus is capable of partial denitrification, and in addition its genome contains the nrcN gene that encodes a second putative NDH-2. Of particular interest is the fact that nrcN is part of an operon (nrcDEFN) that is proposed to encode a protein complex specifically required for nitrate reduction. In this work, the nrcN gene has the activity expected of a NDH-2, and functions independently of other components of the putative Nrc complex. The biochemical properties of the two NDH-2 enzymes are compared. Efforts to demonstrate that NrcN is part of a multiprotein complex were not successful. However, the NrcE protein was expressed in Escherichia coli and shown to be a membrane-bound protein containing heme B.
Mechanistic aspects of sodium-binding sites in LeuT-like fold symporters
Authors: Perez Camilo , Ziegler Christine
Published: February 2013

Secondary active transporters are of paramount biological impact in all living cells, facilitating the movement of many different substrates across the membrane against a concentration gradient. The uphill transport of one substrate is coupled to the downhill transport of another and driven by the electrochemical gradient. In the last decade, an increasing number of atomic structures of secondary transporters have been reported, confirming a very fundamental mechanistic concept known as the alternating-access cycle. The wealth of structures of transporters sharing the so-called LeuT-like fold that is characterized by two five-transmembrane-helix repeats sharing a 2-fold inverted pseudo symmetry has raised big hopes to finally describe alternating access on a molecular level. Although comparing the individual transporter states of different LeuT-like fold transporters revealed striking similarities, the coupling process, which represents the heart of secondary transport, is far from being understood. Here, we review the structural, functional, and biophysical validation of sodium-binding sites in four different LeuT-like fold transporters. The conservation of sodium sites is discussed in light of their role as key elements connecting symmetry-related structural domains, which are involved in substrate translocation. Moreover, we highlight their crucial roles in conformational changes of LeuT-like fold transporters and their implication on a unifying mechanism in secondary transport.
Hydrolysis of dipeptide derivatives reveals the diversity in the M49 family
Authors: Jajčanin-Jozić Nina , Abramić Marija
Published: February 2013

Dipeptidyl peptidase III, a metallopeptidase of the M49 family, was first identified (in the pituitary) by its specific cleavage of diarginyl arylamides, which have been used as preferred assay substrates until now. Here we examined the activity of the yeast and human dipeptidyl peptidase III in parallel. The human enzyme preferred Arg2-β-naphthylamide and showed 620-fold higher k
cat/K
m for this substrate. In contrast, the yeast enzyme did not display a preference for any of the X-Arg-β-naphthylamide analyzed. The replacement of Gly505 with Asp, resulted in a less active, but more selective, yeast enzyme form. These results indicate diversity in cleavage specificity in the M49 family.
Heparan sulfate: a key regulator of embryonic stem cell fate
Authors: Kraushaar Daniel C. , Dalton Stephen , Wang Lianchun
Published: February 2013

Heparan sulfate (HS) belongs to a class of glycosaminoglycans and is a highly sulfated, linear polysaccharide. HS biosynthesis and modification involves numerous enzymes. HS exists as part of glycoproteins named HS proteoglycans, which are expressed abundantly on the cell surface and in the extracellular matrix. HS interacts with numerous proteins, including growth factors, morphogens, and adhesion molecules, and thereby regulates important developmental processes in invertebrates and vertebrates. Embryonic stem cells (ESCs) are distinguished by their characteristics of self-renewal and pluripotency. Self-renewal allows ESCs to proliferate indefinitely in their undifferentiated state, whereas pluripotency implies their capacity to differentiate into the three germ layers and ultimately all cell types of the adult body. Both traits are tightly regulated by numerous cell signaling pathways. Recent studies have highlighted the importance of HS in the modulation of ESC functions, specifically their lineage fate. Here, we review the current advances that have been made in understanding the structural changes of HS during ESC differentiation and in deciphering the molecular mechanisms by which HS modulates cell fate. Finally, we discuss the applications of heparinoids and chemical inhibitors of HS biosynthesis for the manipulation of ESC culture and directed differentiation.
Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant
Authors: Thornton D. A. , Allen D. A. , Huggon A. P. , Picton D. J. , Robinson A. T. , Steadman R. J. , Serén T. , Lipponen M. , Kekki T.
Published: February 2013

Specimens for the Sizewell B reactor pressure vessel (RPV) in-service steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained.
Exposure Conditions of Reactor Internals of Rovno VVER-440 Nuclear Power Plant Units 1 and 2
Authors: Grytsenko O. V. , Pugach S. M. , Diemokhin V. L. , Bukanov V. N. , Marek M. , Vandlik S.
Published: February 2013

Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Institute for Nuclear Research Kyiv, Ukraine, and Nuclear Research Institute Rez, Czech Republic, are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Institute for Nuclear Research and at Nuclear Research Institute is shown.
2010 Review of Neutron and Non-Neutron Nuclear Data
Authors: Holden N. E.
Published: February 2013

The results of a review and evaluation of neutron and non-neutron nuclear data published in the scientific literature over the past three years are presented. The status of new chemical elements is examined. Ten elements have had their atomic weight and uncertainty replaced by interval values of upper and lower bounds. Data on revised values for the isotopic composition of the elements are reviewed and new recommended values are presented for germanium. Radioactive half-lives are reviewed and latest values presented which include measurements on nuclides of interest and very long-lived nuclides such as double beta decay, double electron capture, long-lived alpha decay, and long-lived beta decay. The latest information and the status on the evaluation of atomic masses are discussed. Data from new measurements on the very heavy elements (trans-meitnerium elements) are discussed and tabulated. Data on various recent neutron cross-section and resonance integral measurements are also discussed and the latest measurements are tabulated in both cases. The JENDL-4.0 and ENDF=B-VII.1 nuclear data libraries are discussed. A new initiative on the existence and importance of isotopes is presented.
Path Forward for Dosimetry Cross Sections
Authors: Griffin P. J. , Peters C. D.
Published: February 2013

In the 1980's the dosimetry community embraced the need for a high fidelity quantification of uncertainty in nuclear data used for dosimetry applications. This led to the adoption of energy-dependent covariance matrices as the accepted manner of quantifying the uncertainty data. The trend for the dosimetry community to require high fidelity treatment of uncertainty estimates has continued to the current time where requirements on nuclear data are codified in standards such as ASTM E 1018. This paper surveys the current state of the dosimetry cross sections and investigates the quality of the current dosimetry cross section evaluations by examining calculated-to-experimental ratios in neutron benchmark fields. In recent years more nuclear-related technical areas are placing an emphasis on uncertainty quantification. With the availability of model-based cross sections and covariance matrices produced by nuclear data codes, some nuclear-related communities are considering the role these covariance matrices should play. While funding within the dosimetry community for cross section evaluations has been very meager, other areas, such as the solar-related astrophysics community and the US Nuclear Criticality Safety Program, have been supporting research in the area of neutron cross sections. The Cross Section Evaluation Working Group (CSEWG) is responsible for the creation and maintenance of the ENDF/B library which has been the mainstay for the reactor dosimetry community. Given the new trends in cross section evaluations, this paper explores the path forward for the US nuclear reactor dosimetry community and its use of the ENDF/B cross-sections. The major concern is maintenance of the sufficiency and accuracy of the uncertainty estimate when used for dosimetry applications. The two major areas of deficiency in the proposed ENDF/B approach are: 1) the use of unrelated covariance matrices in ENDF/B evaluations and 2) the lack of “due consideration” of experimental data in the evaluation.
New Work on Updating and Extending the Nuclear Data Standards
Authors: Carlson A. D. , Pronyaev V. G. , Capote R. , Hambsch F. -J. , Käppeler F. , Lederer C. , Mannhart W. , Mengoni A. , Nelson R. O. , Schillebeeckx P. , Talou P. , Tagesen S. , Vonach H. , Vorobyev A. , Wallner A.
Published: February 2013

An International Atomic Energy Agency (IAEA) Data Development Project was initiated to provide a mechanism for allowing new experimental data and improvements in evaluation procedures to be incorporated into new evaluations of the standards. The work on this project is ongoing. In the past very long periods sometimes occurred between evaluations of the standards. Through the use of this Project, such long periods should not occur. Work on the cross section standards through this Project has included an update of the experimental data to be used in the cross section standards evaluations, a study of the uncertainties obtained in the international standards evaluation, and improvements in the smoothing procedure for capture cross sections. It was decided that this Project should have a broader range of activities than just the cross section standards and thus encompass standards related activities. The following are being investigated: improvements in the gold cross section at energies below where it is considered a standard and work on certain cross sections that are not as well known as the cross section standards but could be very useful as reference cross sections relative to which certain types of cross section measurements can be made. This work includes prompt gamma-ray production in fast neutron-induced reactions, and work on the 252Cf spontaneous fission neutron spectrum and the 235U thermal neutron fission spectrum. Most of the data investigated through this Project are used in dosimetry applications.
A New Formulation of the Unified Monte Carlo Approach (UMC-B) and Cross-Section Evaluation for the Dosimetry Reaction 55Mn(n,γ)56Mn
Authors: Capote Roberto , Smith Donald L. , Trkov Andrej , Meghzifene Mehdi
Published: February 2013

Two relatively new approaches to neutron cross section data evaluation are described. They are known collectively as Unified Monte Carlo (versions UMC-G and UMC-B). Comparisons are made between these two methods, as well as with the well-known generalized least-squares (GLSQ) technique, through the use of simple, hypothetical (toy) examples. These new Monte Carlo methods are based on stochastic sampling of probability functions that are constructed with the use of theoretical and experimental data by applying the principle of maximum entropy. No further assumptions are involved in either UMC-G or UMC-B. However, the GLSQ procedure requires the linearization of non-linear terms, such as those that occur when cross section ratio data are included in an evaluation. It is shown that these two stochastic techniques yield results that agree well with each other, and with the GLSQ method, when linear data are involved, or when the perturbations due to data discrepancies and nonlinearity effects are small. Otherwise, there can be noticeable differences. The present investigation also demonstrates, as observed in earlier work, that the least-squares approach breaks down when these conditions are not satisfied. This paper also presents an actual evaluation of the 55Mn(n,γ)56Mn neutron dosimetry reaction cross section in the energy range from 100 keV to 20 MeV, which was performed using both GLSQ and UMC-G approaches.
Uncertainty-Accounted Calculational–Experimental Approach for Improved Conservative Evaluations of VVER RPV Radiation Loading Parameters
Authors: Borodkin P. G. , Borodkin G. I. , Khrennikov N. N.
Published: February 2013

The approach of improved uncertainty-accounted conservative evaluation of vodo-vodyanoi energetichesky reactor (VVER) (reactor-) pressure-vessel (RPV) radiation loading parameters has been proposed. This approach is based on the calculational–experimental procedure, which takes into account C/E ratio, depending on over- or underestimation, and uncertainties of measured and calculated results. An application of elaborated approach to the full-scale ex-vessel neutron dosimetry experiments on Russian VVERs combined with neutron-transport calculations has been demonstrated in the paper.
Application of Different Nuclides in Retrospective Dosimetry
Authors: Konheiser J. , Mittag S. , Viehrig H. -W. , Gleisberg B.
Published: February 2013

The activities of nuclides produced via the neutron irradiation of reactor pressure vessel (RPV) steel are used to validate respective fluence calculations. Niobium, nickel, and technetium isotopes from RPV trepans of the decommissioned NPP Greifswald (VVER-440) have been analyzed. The activities were determined by TRAMO (Monte-Carlo) fluence calculations, newly applying 640 neutron-energy groups and ENDF/B7 data. Relative to earlier results, fluences up to 20 % higher have been computed, leading to somewhat better agreement between measurement and calculation, particularly in the case of Tc-99.
Neutron Flux Reduction Programs for Reactor Pressure Vessel of Korea Nuclear Unit 1
Authors: Yoo Choon Sung , Kim Byoung Chul
Published: February 2013

The objective of this work is to implement various fast neutron flux reduction programs on the beltline region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RTPTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RTPTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RTPTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RTPTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library.
A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors
Authors: Morgan D. O. , Robinson A. T. , Allen D. A. , Picton D. J. , Thornton D. A. , Shaw S. E.
Published: February 2013

This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition.
Dosimetry of the Decommissioned High-Flux Beam Reactor at Brookhaven Lab
Authors: Hu J. -P. , Reciniello R. N. , Holden N. E.
Published: February 2013

The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996–2009 is described in the paper.
Benchmarking of Neutron Production of Heavy-Ion Transport Codes
Authors: Remec I. , Ronningen R. M. , Heilbronn L.
Published: February 2013

Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required.
Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods
Authors: Jones Eric N.
Published: February 2013

U.S. Regulatory Guide 1.99 Revision 2 (U.S. Nuclear Regulatory Commission, 1988, “Radiation Embrittlement of Reactor Vessel Materials,” Regulatory Guide 1.99, Revision 2, Washington, D.C.) provides for the use of two substantially different methods for determining through-wall fluence in nuclear reactor pressure vessels. One method is a generic attenuation curve based on a simplistic exponential decay equation. Partly due to the simplicity of its application, the generic attenuation method is predominantly used for licensing calculations. However, it has a limitation in that at increasing distances away from the core beltline, it becomes increasingly less accurate because it cannot account for neutron streaming effects in the cavity region surrounding the pressure vessel. The other attenuation method is based on a displacement per atom (dpa) calculation specific to the reactor vessel structure. The dpa method provides a more accurate representation of fluence attenuation through the reactor pressure vessel (RPV) wall at all elevations of the pressure vessel because it does account for neutron streaming in the cavity region. A requirement for using the dpa method, however, is an accurate flux solution through the RPV wall. This requirement has limited the use of traditional transport methods, such as discrete ordinates, that are limited by their treatment of cavity regions (i.e., air) outside the pressure vessel wall. TransWare Enterprises, under the sponsorship of EPRI and BWRVIP, has developed an advanced three-dimensional transport methodology capable of producing fully converged flux solutions throughout the entire reactor system, including in the cavity region and primary shield structures. This methodology provides an accurate and reliable determination of through-wall fluence in boiling water reactor (BWR) and pressurized water reactor (PWR) pressure vessels, thus allowing the dpa method to be implemented with high reliability. Using this advanced 3-D methodology, this paper presents comparisons of the generic and dpa attenuation methods at critical locations in both BWR and PWR pressure vessel walls.
The Power Distribution and Neutron Fluence Measurements and Calculations in the VVER-1000 Mock-Up on the LR-0 Research Reactor
Authors: Košt'ál M. , Juříček V. , Rypar V. , Švadlenková M. , Cvachovec F.
Published: February 2013

The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed.
Jules Horowitz Reactor, a New Irradiation Facility: Improving Dosimetry for the Future of Nuclear Experimentation
Authors: Grégoire Gilles , Beretz Daniel , Destouches Christophe
Published: February 2013

The Jules Horowitz Reactor (JHR) is an experimental reactor under construction at the French Nuclear Energy and Alternative Energies Commission (CEA) facility at Cadarache. It will achieve its first criticality by the end of 2014. Experiments that will be conducted at JHR will deal with fuel, cladding, and material behavior. The JHR will also produce medical radio-isotopes and doped silicon for the electronic industry. As a new irradiation facility, its instrumentation will benefit from recent improvements. Nuclear instrumentation will include reactor dosimetry, as it is a reference technique to determine neutron fluence in experimental devices or characterize irradiation locations. Reactor dosimetry has been improved with the progress of simulation tools and nuclear data, but at the same time the customer needs have increased: Experimental results must have reduced and assessed uncertainties. This is now a necessary condition to perform an experimental irradiation in a test reactor. Items improved, in the framework of a general upgrading of the dosimetry process based on uncertainty minimization, will include dosimeter, nuclear data, and modelling scheme.
Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G
Authors: Chen J. , Alpan F. A. , Fischer G. A. , Fero A. H.
Published: February 2013

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.
A Broad-Group Cross-Section Library Based on ENDF/B-VII.0 for Fast Neutron Dosimetry Using the CPXSD Methodology
Authors: Alpan F. A.
Published: February 2013

A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF), in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and 58Ni(n,γ) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contributon and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within ± 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190.
Application of Ex-Vessel Neutron Dosimetry Combined with In-core Measurements for Correction of Neutron Source Used for Reactor Pressure Vessel Fluence Calculations
Authors: Borodkin P. G. , Borodkin G. I. , Khrennikov N. N. , Konheiser J.
Published: February 2013

This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data.
The Fast Neutron Fluence and the Activation Detector Activity Calculations Using the Effective Source Method and the Adjoint Function
Authors: Hep J. , Konečná A. , Krýsl V. , Smutný V.
Published: February 2013

This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with fewgroup method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reactor. Both of calculation methods are briefly described and the results are compared.
CALMOS: Innovative Device for the Measurement of Nuclear Heating in Material Testing Reactors
Authors: Carcreff H.
Published: February 2013

An R&D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Division at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2 W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed.
Development of INSPCT-S for Inspection of Spent Fuel Pool
Authors: Walters William , Haghighat Alireza , Sitaraman Shivakumar , Ham Young
Published: February 2013

In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, (α, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data.
An Alternative Calibration Method for Counting P-32 Reactor Monitors
Authors: Quirk T. J. , Vehar D. W.
Published: February 2013

Radioactivation of sulfur is a common technique used to measure fast neutron fluences in test and research reactors. Elemental sulfur can be pressed into pellets and used as monitors. The 32S(n, p)32P reaction has a practical threshold of about 3 MeV and its cross section and associated uncertainties are well characterized [1]. The product 32P emits a beta particle with a maximum energy of 1710 keV [2]. This energetic beta particle allows pellets to be counted intact. ASTM Standard Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32 (E265) [3] details a method of calibration for counting systems and subsequent analysis of results. This method requires irradiation of sulfur monitors in a fast-neutron field whose spectrum and intensity are well known. The resultant decay-corrected count rate is then correlated to the known fast neutron fluence. The Radiation Metrology Laboratory (RML) at Sandia has traditionally performed calibration irradiations of sulfur pellets using the 252Cf spontaneous fission neutron source at the National Institute of Standards and Technology (NIST) [4] as a transfer standard. However, decay has reduced the intensity of NIST's source; thus lowering the practical upper limits of available fluence. As of May 2010, neutron emission rates have decayed to approximately 3e8 n/s. In practice, this degradation of capabilities precludes calibrations at the highest fluence levels produced at test reactors and limits the useful range of count rates that can be measured. Furthermore, the reduced availability of replacement 252Cf threatens the long-term viability of the NIST 252Cf facility for sulfur pellet calibrations. In lieu of correlating count rate to neutron fluence in a reference field the total quantity of 32P produced in a pellet can be determined by absolute counting methods. This offers an attractive alternative to extended 252Cf exposures because it can be performed regardless of the characterization of the exposure environment. Count rates produced by sulfur pellets are correlated to the measured quantity of separated 32P. A posteriori spectral and cross section determination can be used to correlate the quantity of phosphorus back to a neutron fluence in a reference field. This paper outlines a method for the setup, calibration, and use of the detector systems, 32P sample preparation, and analysis of the beta spectrum. An uncertainty analysis and comparison to ASTM E265 is also included.
New Experimental Proposal for 235U PFNS Measurement to Answer a Fifty Year Old Question
Authors: Kornilov N. , Massey T. , Grimes S. , Voinov A.
Published: February 2013

The Prompt Fission Neutron Spectrum (PFNS) from 235U(n,f) is very important for various nuclear applications. It has been investigated in different experiments. In spite of ~ 50 years of experimental efforts, a continuing conflict exists at thermal neutron energy. Microscopic experimental PFNS cannot describe macroscopic data. In this report we discuss the current status of this problem and suggest a new experiment, which could possibly resolve this problem.
Development of Neutron Measurement in Intense Gamma Field Using New Type of Nuclear Emulsion
Authors: Kawarabayashi J. , Ishihara K. , Takagi K. , Tomita H. , Iguchi T. , Naka T. , Morishima K. , Maeda S.
Published: February 2013

To precisely measure the neutron emissions from a spent fuel assembly of a fast breeder reactor, we formed nuclear emulsions based on a non-sensitized Oscillation Project with Emulsion tRacking Apparatus (OPERA) film with AgBr grain sizes of 60, 90, and 160 nm. The efficiency for 252Cf neutron detection of the new emulsion was calculated to be 0.7 × 10−4, which corresponded to an energy range from 0.3 to 2 MeV and was consistent with a preliminary estimate based on experimental results. The sensitivity of the new emulsion was also experimentally estimated by irradiating with 565 keV and 14 MeV neutrons. The emulsion with an AgBr grain size of 60 nm had the lowest sensitivity among the above three emulsions but was still sensitive enough to detect protons. Furthermore, the experimental data suggested that there was a threshold linear energy transfer of 15 keV/μm for the new emulsion, below which no silver clusters developed. Further development of nuclear emulsion with an AgBr grain size of a few tens of nanometers will be the next stage of the present study.
Crystal structure of μ3-fluoro-tri-μ2-fluoro-tris(tetramethylethylenediamine nickel(II)) fluoride trihydrate, [Ni3(tmen)3F5]+·F–·3H2O, C18H54F6N6Ni3O3
Authors: Ishikawa Yuki , Miyamoto Keiko , Horn Ernst
Published: February 2013

C18H54F6N6O3Ni3, triclinic, P1̄ (no. 2), a = 8.6908(7) Å, b = 13.717(1) Å, c = 14.281(1) Å, α = 72.997(1)°, β = 89.459(1)°, γ = 75.718(1)°, V = 1574.1(2) Å3, Z = 2, Rgt(F) = 0.0355, wRref(F2) = 0.0433, T = 273,2 K.
“Doing Ô Sin”
Authors: Minh Nguyễn Thị Nguyệt
Published: February 2013

This article discusses the negotiations of migrant domestic workers in Hà Nội with the urban labor market, the conditions and relations of paid domestic work, and their life away from home. These negotiations indicate constraining structures of the labor market and the challenging conditions of urban life and work. They shape and are shaped by the power relations that the domestic workers simultaneously resist and legitimize through an active investment in the cultural discourse of reciprocal affection and harmony. The negotiations, however, demonstrate the diverse ways in which domestic workers actively carve out niches of autonomy and control over their life and work.
Migration, Social Networks and Risk
Authors: Winkels Alexandra
Published: February 2013

This study of rural-to-rural migration in Vietnam investigates the role of social networks in the migration process and how they affect the risks associated with mobility. The findings point to a complex picture in which migrants use their social capital to reduce the challenges associated with moving, settling and both accessing and maintaining income opportunities at the destination. At the same time, the pooling of financial and human resources at the origin household level allows migrants to take greater economic risks at the destination, thus potentially increasing levels of livelihood vulnerability for both the migrant and those household members left behind.
Liquid meal composition, postprandial satiety hormones, and perceived appetite and satiety in obese women during acute caloric restriction
Authors: Heden Timothy D , Liu Ying , Sims Lauren , Kearney Monica L , Whaley-Connell Adam T , Chockalingam Anand , Dellsperger Kevin C , Fairchild Timothy J , Kanaley Jill A
Published: February 2013

Objective

The purpose of this study was to compare postprandial satiety regulating hormone responses (pancreatic polypeptide (PP) and peptide tyrosine tyrosine (PYY)) and visual analog scale- (VAS) assessed perceived appetite and satiety between liquid high-protein (HP) and high-carbohydrate (HC) meals in obese women during acute (24-h) caloric restriction.

Design

Eleven obese premenopausal women completed two conditions in random order in which they consumed 1500 calories as six 250-calorie HP meals or six 250-calorie HC meals over a 12-h period. Blood samples were taken at baseline and every 20 min thereafter and analyzed for PP and PYY concentrations. At these same points, perceived hunger and fullness were assessed with a VAS. The incremental area under the curve (iAUC) was used to compare postprandial responses.

Results

The 12-h PP and PYYiAUC were greater (P≤0.05) during the HP condition (PP: 4727±1306 pg/ml×12 h, PYY: 1373±357 pg/ml×12 h) compared with the HC condition (PP: 2300±528 pg/ml×12 h, PYY: 754±246 pg/ml×12 h). Perceived hunger and fullness were not different between conditions (P>0.05). The greatest changes in PYY and perceived fullness occurred after the morning meals during both conditions.

Conclusions

These data suggest that in obese women during acute caloric restriction before weight loss, i) liquid HP meals, compared with HC meals, result in greater postprandial PP and PYY concentrations, an effect not associated with differential appetite or satiety responses, and ii) meal-induced changes in PYY and satiety are greatest during the morning period, regardless of dietary macronutrient composition.
The role of the renin–angiotensin–aldosterone system in preeclampsia: genetic polymorphisms and microRNA
Authors: Yang Jie , Shang Jianyu , Zhang Suli , Li Hao , Liu Huirong
Published: February 2013

The compensatory alterations in the rennin–angiotensin–aldosterone system (RAAS) contribute to the salt–water balance and sufficient placental perfusion for the subsequent well-being of the mother and fetus during normal pregnancy and is characterized by an increase in almost all the components of RAAS. Preeclampsia, however, breaks homeostasis and leads to a disturbance of this delicate equilibrium in RAAS both for circulation and the uteroplacental unit. Despite being a major cause for maternal and neonatal morbidity and mortality, the pathogenesis of preeclampsia remains elusive, where RAAS has been long considered to be involved. Epidemiological studies have indicated that preeclampsia is a multifactorial disease with a strong familial predisposition regardless of variations in ethnic, socioeconomic, and geographic features. The heritable allelic variations, especially the genetic polymorphisms in RAAS, could be the foundation for the genetics of preeclampsia and hence are related to the development of preeclampsia. Furthermore, at a posttranscriptional level, miRNA can interact with the targeted site within the 3′-UTR of the RAAS gene and thereby might participate in the regulation of RAAS and the pathology of preeclampsia. In this review, we discuss the recent achievements of genetic polymorphisms, as well as the interactions between maternal and fetal genotypes, and miRNA posttranscriptional regulation associated with RAAS in preeclampsia. The results are controversial but utterly inspiring and attractive in terms of potential prognostic significance. Although many studies suggest positive associations with genetic mutations and increased risk for preeclampsia, more meticulously designed large-scale investigations are needed to avoid the interference from different variations.