bluuebunny/crossref_metadata_2025_split huggingface dataset:
The genetic polymorphisms of HER-2 and the risk of lung cancer in a Korean population
Authors: Jo Uk Hyun , Han Sle Gi Lo , Seo Jae Hong , Park Kyong Hwa , Lee Jae Won , Lee Hyo Jung , Ryu Jeong Seon , Kim Yeul Hong 
Published: December 2008

Abstract

Background
Human Epidermal Growth Factor Receptor 2 (HER-2; also known as erbB-2 or neu), a proto-oncogene of the receptor tyrosine kinase superfamily, has been associated with carcinogenesis and prognosis of human cancers, acting as a binding partner of other epidermal growth factor receptor (EGFR) family in the activation of EGFR signaling. Amplification of the HER-2 gene has been reported in lung cancer, where it has been associated with poor prognosis. In this study, we investigated whether the four polymorphisms (-3444C>T, -1985 G>T, I655A A>G and P1170A C>G) of the HER-2 gene are associated with the risk of lung cancer in Korean populations.

Methods
The frequencies of 4 polymorphisms of the HER-2 gene were examined by the polymerase chain reaction-restriction fragment length polymorphism or the single-nucleotide polymorphism-identification technology assay in the 407 lung cancer patients and 407 healthy controls.

Results
The frequencies of the 4 polymorphisms were not significantly different between patient and control groups in overall subjects. However, in the subgroup analysis, the 3 single nucleotide polymorphisms (-3444C>T, -1985G>T and P1170A C>G) showed statistically significant differences in the subgroups of females, non-smokers, and non-drinkers (p < 0.05). Additionally, we found the association between the risk of lung cancer and the polymorphisms of HER-2 gene in non-smoker subgroups with adenocarcinoma (p < 0.05).

Conclusion
Our results suggest that the polymorphisms of the HER-2 gene are associated with an increased susceptibility to lung cancer in females, non-smokers and non-drinkers subgroups in the Korean population.
The association between diabetes mellitus, glucose, and chronic musculoskeletal complaints. Results from the Nord-Trøndelag Health Study
Authors: Hoff Ole M , Midthjell Kristian , Zwart John-Anker , Hagen Knut 
Published: December 2008

Abstract

Background
The relationship between diabetes mellitus (DM) and chronic musculoskeletal complaints (MSCs) is unclear. The aim of this study was to investigate the association between DM, non-fasting glucose and chronic MSCs defined as pain and/or stiffness ≥ 3 months during the past year in the general adult population.

Methods
The results were based on cross-sectional data from 64,785 men and women (aged ≥ 20 years) who participated in the Nord-Trøndelag Health Survey, which included 1,940 individuals with known DM. Associations were assessed using multiple logistic regression, estimating prevalence odds ratio (OR) with 95% confidence intervals (CIs).

Results
High non-fasting glucose was associated with a lower prevalence of chronic MSCs compared to a low glucose level. DM was associated with higher prevalence of chronic MSCs, in particular chronic widespread MSCs. In the multivariate analysis, adjusting for glucose level, BMI, age, gender and physical activity, chronic widespread MSCs was 1.6 times more likely (OR = 1.6, 95% CI 1.2–2.2) among individuals < 60 years of age with DM than among those without DM. The association between chronic widespread MSCs and DM was most evident among the group of individuals aged < 60 years with either type 2 DM or unclassified DM (OR = 1.8, 95% CI 1.3–2.5).

Conclusion
In this cross-sectional study a high non-fasting glucose was associated with lower prevalence of chronic MSCs. Among individuals with known DM chronic widespread MSCs were more likely.
Metabolic engineering of Saccharomyces cerevisiae for the production of n-butanol
Authors: Steen Eric J , Chan Rossana , Prasad Nilu , Myers Samuel , Petzold Christopher J , Redding Alyssa , Ouellet Mario , Keasling Jay D 
Published: December 2008

Abstract

Background
Increasing energy costs and environmental concerns have motivated engineering microbes for the production of "second generation" biofuels that have better properties than ethanol.

Results and conclusion

Saccharomyces cerevisiae was engineered with an n-butanol biosynthetic pathway, in which isozymes from a number of different organisms (S. cerevisiae, Escherichia coli, Clostridium beijerinckii, and Ralstonia eutropha) were substituted for the Clostridial enzymes and their effect on n-butanol production was compared. By choosing the appropriate isozymes, we were able to improve production of n-butanol ten-fold to 2.5 mg/L. The most productive strains harbored the C. beijerinckii 3-hydroxybutyryl-CoA dehydrogenase, which uses NADH as a co-factor, rather than the R. eutropha isozyme, which uses NADPH, and the acetoacetyl-CoA transferase from S. cerevisiae or E. coli rather than that from R. eutropha. Surprisingly, expression of the genes encoding the butyryl-CoA dehydrogenase from C. beijerinckii (bcd and etfAB) did not improve butanol production significantly as previously reported in E. coli. Using metabolite analysis, we were able to determine which steps in the n-butanol biosynthetic pathway were the most problematic and ripe for future improvement.
Myxoma viral serpin, Serp-1, inhibits human monocyte adhesion through regulation of actin-binding protein filamin B
Authors: Viswanathan Kasinath , Richardson Jakob , Togonu-Bickersteth Babajide , Dai Erbin , Liu Liying , Vatsya Pracha , Sun Yun-ming , Yu Jeff , Munuswamy-Ramanujam Ganesh , Baker Henry , Lucas Alexandra R 
Published: December 2008

AbstractSerp-1 is a secreted myxoma viral serine protease inhibitor (serpin) with proven, highly effective, anti-inflammatory defensive activity during host cell infection, as well as potent immunomodulatory activity in a wide range of animal disease models. Serp-1 binds urokinase-type plasminogen activator (uPA) and the tissue-type PA, plasmin, and factor Xa, requiring uPA receptor (uPAR) for anti-inflammatory activity. To define Serp-1-mediated effects on inflammatory cell activation, we examined the association of Serp-1 with monocytes and T cells, effects on cellular migration, and the role of uPAR-linked integrins and actin-binding proteins in Serp-1 cellular responses. Our results show that Serp-1 associates directly with activated monocytes and T lymphocytes, in part through interaction with uPAR (P&lt;0.001). Serp-1, but not mammalian serpin PA inhibitor-1 (PAI-1), attenuated cellular adhesion to the extracellular matrix. Serp-1 and PAI-1 reduced human monocyte and T cell adhesion (P&lt;0.001) and migration across endothelial monolayers in vitro (P&lt;0.001) and into mouse ascites in vivo (P&lt;0.001). Serp-1 and an inactive Serp-1 mutant Serp-1(SAA) bound equally to human monocytes and T cells, but a highly proinflammatory mutant, Serp-1(Ala6), bound less well to monocytes. Serp-1 treatment of monocytes increased expression of filamin B actin-binding protein and reduced CD18 (β-integrin) expression (P&lt;0.001) in a uPAR-dependent response. Filamin colocalized and coimmunoprecipitated with uPAR, and short interference RNA knock-down of filamin blocked Serp-1 inhibition of monocyte adhesion. We report here that the highly potent, anti-inflammatory activity of Serp-1 is mediated through modification of uPAR-linked β-integrin and filamin in monocytes, identifying this interaction as a central regulatory axis for inflammation.
Evaluating the Onset of Tearing in Elastic-Plastic Fracture Toughness Testing Using In Situ Optical Microscopy
Authors: Melcher Ryan J. 
Published: December 2008

Abstract
Fracture toughness, in the sense of material resistance to ductile tearing from an initial sharp defect, is a common metric for structural integrity assessments of engineering components. While standardized test methods are well-suited for repeatable estimation of this metric, physical observation of crack-tip opening displacement (CTOD) and tearing may provide a supplemental means of evaluating the onset of tearing with greater accuracy in ductile materials. In contrast to previously documented methods of physical CTOD measurement, in situ optical microscopy on standard sidegrooved fracture toughness specimens presents an easily implemented, cost-effective tool for observing tearing onset. As an additional benefit, quantitative measurements of CTOD from in situ optical microscopy also provide a means of cross-checking standard J-integral results as determined from load-displacement test data.
Reactor Dosimetry and Nuclear Reactor Regulation
Authors: Lois L 
Published: December 2008

An important function of reactor dosimetry is the verification of fluence values used in the calculation of pressure vessel embrittlement and its impact on (a) the fracture toughness requirements for protection against pressurized thermal shock (Title 10, Code of Federal Regulations [CFR], Part 50.61) and (b) the pressure-temperature operating limits required by 10 CFR 50.60 and Appendix G. The dosimetry surveillance requirements are part of Appendix H to 10 CFR 50.
In recent years there has been significant progress in the computational techniques of neutron transport, making it possible to derive accurate estimates of the fast neutron fluence through the reactor vessel thickness and in the reactor cavity. In addition, a trend has been identified in the material properties and chemical composition of the pressure vessel of recent vintage plants where the materials used are such that the estimated end-of-life fluence is only a very small fraction of the fluence required to meet the 10 CFR 50.61 criteria. For future light water reactors (LWRs), cavity dosimetry combined with analytical computations could give an accurate estimate of the pressure vessel neutron irradiation. This could then be used to estimate the degree of vessel material embrittlement and thus alleviate the need for invessel dosimetry surveillance. The question as to whether the need for archival sample irradiation can be alleviated in the future remains to be answered.
The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP): Past Accomplishments, Recent Developments, and Future Directions
Authors: Gold R , McElroy WN 
Published: December 2008

The integrity of the pressure vessel (PV) of a nuclear power plant is a major safety consideration throughout the life of the power plant. Radiation embrittlement can limit the serviceable lifetime of the PV and thereby limit the effective operating lifetime of the plant. It can further restrict normal heat-up and cool-down reactor operations, with resultant cycle-to-cycle economic implications. In recognition of these safety and economic issues, the U.S. Nuclear Regulatory Commission (NRC) established the Light Water Reactor Pressure Vessel (LWR-PV) Surveillance Dosimetry Improvement Program (SDIP) some ten years ago to improve, maintain, and standardize neutron dosimetry, damage correlation, and the associated reactor analysis procedures used for predicting the integrated effect of neutron exposure on LWR-PV.
The LWR-PV-SDIP adopted specific experimental and calculational strategies to meet the challenge of this complex radiation-induced PV embrittlement phenomenon. A vigorous research effort has gone forward worldwide to implement these strategies. The major benefit of this program has been and continues to be a significant improvement in the accuracy of the assessment of the current metallurgical condition and the remaining safe operating lifetime of LWR-PV.
The LWR-PV-SDIP has produced a broad range of technical accomplishments over the last decade and these achievements are reviewed. These earlier LWR-PV-SDIP accomplish-ments have generated, in turn, a number of significant new developments that are described. A natural outgrowth of LWR-PV-SDIP work is the experience to project future needs of LWR-PV surveillance. On this basis, recommendations for future directions are advanced with special emphasis on plant life extension.
Analysis of First Pressure Vessel Surveillance Capsule from Kršsko Nuclear Power Plant
Authors: Kodeli I , Remec I , Glumac B , Najžer M , Loose A 
Published: December 2008

Results of the analysis of the first pressure vessel (PV) radiation embrittlement surveillance capsule from Krško nuclear power plant are presented. The capsule was removed and tested after 1.75 effective full power years of operation. Neutron flux and spectrum were calculated with DOT 3.5 and 4.2 transport codes at the locations of the specimens in the capsule and in the pressure vessel. Nickel, iron, copper, cobalt, Np-237, and U-238 neutron dosimeters were irradiated in surveillance capsules. Experimentally determined reaction rates were compared to values calculated from the neutron transport calculation,with source strength corresponding to the nominal reactor thermal power. Good agreement (within 15%) was found for threshold neutron detectors. The STAY'SL unfolding code was used to adjust the measured and calculated reaction rates and neutron fluences.
Shifts in the RTNDT temperature determined from Charpy V-notch curves for irradiated material relative to that of unirradiated materials were compared with shifts predicted by NRC Regulatory Guide 1.99, Rev. 1 and Rev. 2, at the neutron exposure of 7 to 9 E18 n∙cm−2 (ϕ(E &gt; 1 MeV)) experienced by specimens taking into account the chemical composition of PV steel.
Application of Ex-Vessel Neutron Dosimetry for Determination of Vessel Fluence
Authors: Lippincott EP , Anderson SL , Fero AH 
Published: December 2008

Routine reactor cavity dosimetry measurements in four plants have been completed. Each of these measurements has included several multiple reaction sets that contain radiometric monitors and solid-state track recorders (SSTRs). The SSTRs are used to accurately determine fission rates using only extremely small amounts of very high purity fissionable material. The multiple reaction dosimetry sets were contained in holders suspended by wires extending from the top to the bottom of the cavity. These stainless steel wires were also analyzed to provide both fast and thermal neutron axial gradient data.
A detailed discrete ordinates transport theory calculation was performed for each plant. These calculations included the cycle average neutron source distribution in the fuel and a detailed model of the cavity region. The calculations in general reproduce the measured gradient shapes but exhibit a consistent absolute bias. Although part of this bias probably arises from known cross-section deficiencies, some of the bias is estimated to arise from other, unknown, sources. Because of the general consistent bias of the measurements, a simple renormalization of the calculation allows the neutron flux to be evaluated at all points in the reactor vessel.
Power Reactor Neutron Dosimetry: Experimental Estimate of the Influences Due to Local Changes in Core Configuration and Gamma-Induced Reactions
Authors: Alloggio G , Borioli E , Brega E , Sandrelli G , Cesana A , Terrani M 
Published: December 2008

For the evaluation of neutron flux densities in reactor pressure vessel surveillance analyses a detailed knowledge of the local power history is required. In this paper a new method is proposed which reconstructs experimentally the effects of the local power history through the analysis of the measured activities in fission dosimeters. The validity of the method was tested by comparing measurements carried out at Caorso BWR plant with neutron flux density computations performed using the DOT 4.2 code.
The same measurements and computations were utilized for carrying out an estimate of the gamma-ray flux density at the irradiation position.
Review of the NESTOR Shielding and Dosimetry Improvement Programme (NESDIP)
Authors: Butler J , Carter MD , Curl IJ , Miller PC , Packwood A , Power S 
Published: December 2008

The NESTOR Shielding and Dosimetry Improvement Programme (NESDIP) was designed to study neutron penetration through typical PWR radial shields. The programme also encompasses neutron transport within the reactor cavity.
The NESDIP experiments are conducted in the ASPIS shielding facility on the NESTOR reactor at AEE Winfrith. An essential feature of these experiments is the use of a fission plate to provide a well-characterized neutron source.
The development of the experimental programme is described. It commenced with a replica of the Oak Ridge PCA radial shield benchmark, which was followed by the ASPIS-PCA slab geometry benchmarks and the cavity/nozzle and coolant-duct benchmarks. The role of future NESDIP experiments is examined.
The calculational scheme developed for PWR shield design at Winfrith involves a linked sequence of codes. Point-Kernel, Kernel-Albedo, Diffusion and Monte Carlo methods are employed for individual stages of the calculation. The choice of method at each stage is made with regard to the specific physics and geometric characteristics of that stage.
The application of the calculational route for LWR cavity streaming is described and an assessment of the accuracies made by analysis of the NESDIP measurements.
Power Reactor Benchmark Studies
Authors: Lippincott EP , McElroy WN 
Published: December 2008

The Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) was established by the U.S. Nuclear Regulatory Commission (NRC) to improve, test, verify, and standardize the physics-dosimetry-metallurgy measurements and calculations used to predict the effect of neutron exposure to LWR pressure vessels and their support structures. As part of this effort the PCA and PSF experiments were initiated at Oak Ridge National Laboratory (ORNL) to allow accurate measurements adjacent to and through a pressure vessel mockup.
It was recognized, however, that additional benchmark measurements would be needed to relate the PCA/PSF data to operating plants. Although measurements in operating plants could not be nearly as extensive as those in a mockup, only actual plant measurements can provide a test of the application of calculational techniques to the actual, more complex, LWR geometry. Therefore, in cooperation with vendors, utilities, and other laboratories, dosimetry measurements have been made in several operating plants. These measurements were carried out in the reactor cavities, adjacent to the outside of the reactor vessel, and, in one case, in a special capsule located at a surveillance position inside the vessel.
The reactor measurement results provide a series of data points for validation of transport calculation methods. The value of the data lies in the evaluation of absolute magnitudes of flux through the vessel and in the information on azimuthal and axial flux variations. Application of calculational techniques to evaluation of various flux reduction schemes is of particular concern as regards these variations.
Results from the power reactor benchmarks are generally consistent with bias trends observed in the mockup measurements. It is anticipated that future calculational improvements will utilize the power reactor benchmark data for verification.
Nuclear Research Emulsion Measurements and the Observation of Proportional Counter Perturbation Factors in VENUS-1
Authors: Gold R , Preston CC , Roberts JH , DeLeeuw G , Fabry A , Leenders L 
Published: December 2008

Nuclear research emulsion (NRE) observed I-integral and J-integral reaction rates are reported for VENUS-1. Proton-recoil energy-dependent reaction rates are predicted using CEN/SCK calculated 17 group neutron energy spectra. These calculated reaction rates are compared with absolute NRE observations at seven different locations in steel and water regions of VENUS-1. The agreement between calculations and NRE observed absolute proton-recoil reaction rates depends sensitively on location in VENUS-1. Adequate agreement is attained near the center of the core. However, as one approaches the edge of the core corner at approximately 28°, the agreement deteriorates. Comparisons between theory and experiment are limited by the broad energy resolution of the 17 energy group structure used in the calculations. Nevertheless, these comparisons do suggest that difficulties arise in the calculation of neutron transport as one approaches the edge of the core corner.
In the 1983 dosimetry campaign in VENUS-1, possible perturbations introduced by 4π proton-recoil gas-filled proportional counters were investigated with NRE. Preliminary I-integral and J-integral proton-recoil reaction rates from NRE irradiated in this perturbation study reveal that proportional counters do create non-negligible perturbations in LWR-PV environments.
Neutronic Study of a PAHR Irradiation Experiment in BR2
Authors: Fouarge L , D'hondt P , De Raedt C , Debrue J , Joly C 
Published: December 2008

Post-accident heat removal (PAHR) experiments are being conducted in BR2 as part of the fast reactor safety program of the Commission of the European Communities. In the first experiment a mixture of UO2 particles in a crucible with sodium has been irradiated in the central channel of BR2 at a temperature of about 1500°. Two similar experiments will be performed in the future at higher temperatures.
Neutron and gamma calculations (one- and two-dimensional) contributed to the optimization of the irradiation device and determined the heating rate distributions and level. The latter is compared with the results of an analysis of the data collected during the irradiation. As quality control of the particle bed, a neutron transmission measurement with a beam issuing from the BR1 reactor allowed us, before the irradiation, to scan the bed in order to determine its porosity variation.
Determination of Neutron Exposure in the Buffalo Reactor
Authors: Lippincott EP , Rogers JW , Hawthorne JR 
Published: December 2008

In a paper presented at the 5th ASTM-Euratom Symposium, the determination of neturon flux in the light water test reactor at the Nuclear Science and Technology Facility of the State University of New York at Buffalo was reported. In this experiment, dosimetry measurements were made in a mockup of a configuration used for typical metallurgical irradiations. A detailed calculation was also performed to determine the neutron spectrum for correlation of the measured results.
Further measurements have been made using dosimetry capsules inserted into metallurgical experiments. Since the Buffalo reactor has a very low rate of fuel burnup, it is expected that conditions should remain constant except when changes are occasionally made in fuel element locations in the vicinity of the irradiation experiment. Of more significance, perhaps, are changes in the irradiation rig itself, whether due to changes in composition or rig geometry.
Other effects that are of importance to the evaluation of the fluence received by the metallurgical samples are due to gradient effects within the irradiation block, both axial and radial. No measurement has been made of the radial gradients, but these may be estimated using the transport calculation. Measurements of the axial gradients have indicated that these are within tolerable limits (̃10%) due to the limited axial distance filled with samples. In addition, results from dosimetry in the metallurgical tests have been found to be consistent with the dosimetry test results.
It is concluded that irradiation conditions can be well determined for each sample, including gradient and time variation effects. It is recommended, however, that dosimetry measurements of each experiment be continued to ensure highest confidence in exposure results and to guarantee that effects due to changing conditions (as noted above) are recognized.