The Uses of Argument in Mathematics

Stephen Toulmin once observed that `it has never been customary for philosophers to pay much attention to the rhetoric of mathematical debate'. Might the application of Toulmin's layout of arguments to mathematics remedy this oversight? Toulmin's critics fault the layout as requiring so much abstraction as to permit incompatible reconstructions. Mathematical proofs may indeed be represented by fundamentally distinct layouts. However, cases of genuine conflict characteristically reflect an underlying disagreement about the nature of the proof in question.Comment: 10 pages, 5 figures. To be presented at the Ontario Society for the Study of Argumentation Conference, McMaster University, May 2005 and LOGICA 2005, Hejnice, Czech Republic, June 200

Taking Blockchain Seriously

In the present techno-political moment it is clear that ignoring or dismissing the hype surrounding blockchain is unwise, and certainly for regulatory authorities and governments who must keep a grip on the technology and those promoting it, in order to ensure democratic accountability and regulatory legitimacy within the blockchain ecosystem and beyond. Blockchain is telling (and showing) us something very important about the evolution of capital and neoliberal economic reason, and the likely impact in the near future on forms and patterns of work, social organization, and, crucially, on communities and individuals who lack influence over the technologies and data that increasingly shape and control their lives. In this short essay I introduce some of the problems in the regulation of blockchain and offer counter-narratives aimed at cutting through the hype fuelling the ascendency of this most contemporary of technologies

Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data

In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC) experiments was conducted by the Institut de Radioprotection et de Surete Nucleaire (IRSN) at the experimental criticality facility in Valduc, France. The plutonium-to- uranium ratio and the isotopic compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an enrichment of 4.5 wt% {sup 235}U and was burned to 37,500 MWd/MTU. The fuel material also includes {sup 241}Am, which is present due to the decay of {sup 241}Pu. The HTC experiments include configurations designed to simulate fuel handling activities, pool storage, and transport in casks constructed of thick lead or steel. Rights of use for the HTC experiment data were purchased under an agreement that limits release of the information. Consequently, a detailed and complete description of the experiments is not presented in this report. This report discusses evaluation of the four HTC data reports, modeling of the experiments, sensitivity and uncertainty analysis, and upper subcritical limit (USL) calculation. The report also presents some conclusions and recommendations concerning use of the HTC experiment data for burnup credit applications. The similarity of the HTC experiments with PWR spent nuclear fuel has been quantified using sensitivity/uncertainty analysis, confirming that the HTC experiments are significantly more applicable to the validation of burnup credit calculations than other available mixed-oxide (MOX) experiments. The HTC experiments were designed and executed with a high level of rigor, resulting in experimental uncertainties that are lower than many of the earlier MOX experiments. The HTC data reports, together with information provided in this report, provide sufficient data to allow for either detailed or simplified computational models to be developed. The HTC experiments substantially strengthen the technical basis for validation by adding to the previously small number of applicable experiments against which to compare burnup credit applications. Existing MOX experiments found to be applicable are also identified in this report

Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the international nuclear data community as of March 2005. The accuracy of the cross-section data was investigated by comparing existing cross-section evaluations against available measured cross-section data. When possible, benchmark calculations were also used to assess the performance of the latest FP cross-section data. Since March 2005, the U.S. and European data projects have released newer versions of their respective data files. Although there have been updates to the international data files and to some degree FP data, much of the updates have included nuclear cross-section modeling improvements at energies above the resonance region. The one exception is improved ENDF/B-VII cross-section uncertainty data or covariance data for gadolinium isotopes. In particular, ENDF/B-VII includes improved 155Gd resonance parameter covariance data, but they are based on previously measured resonance data. Although the new covariance data are available for 155Gd, the conclusions of the FP cross-section data assessment of this report still hold in lieu of the newer international cross-section data files. Based on the FP data assessment, there is judged to be a need for new total and capture cross-section measurements and corresponding cross-section evaluations, in a prioritized manner, for the nine FPs to provide the improved information and technical rigor needed for criticality safety analyses

Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister

Performance Evaluation of Supercomputers using HPCC and IMB Benchmarks

The HPC Challenge (HPCC) benchmark suite and the Intel MPI Benchmark (IMB) are used to compare and evaluate the combined performance of processor, memory subsystem and interconnect fabric of five leading supercomputers - SGI Altix BX2, Cray XI, Cray Opteron Cluster, Dell Xeon cluster, and NEC SX-8. These five systems use five different networks (SGI NUMALINK4, Cray network, Myrinet, InfiniBand, and NEC IXS). The complete set of HPCC benchmarks are run on each of these systems. Additionally, we present Intel MPI Benchmarks (IMB) results to study the performance of 11 MPI communication functions on these systems

TSUNAMI Primer: A Primer for Sensitivity/Uncertainty Calculations with SCALE

This primer presents examples in the application of the SCALE/TSUNAMI tools to generate k{sub eff} sensitivity data for one- and three-dimensional models using TSUNAMI-1D and -3D and to examine uncertainties in the computed k{sub eff} values due to uncertainties in the cross-section data used in their calculation. The proper use of unit cell data and need for confirming the appropriate selection of input parameters through direct perturbations are described. The uses of sensitivity and uncertainty data to identify and rank potential sources of computational bias in an application system and TSUNAMI tools for assessment of system similarity using sensitivity and uncertainty criteria are demonstrated. Uses of these criteria in trending analyses to assess computational biases, bias uncertainties, and gap analyses are also described. Additionally, an application of the data adjustment tool TSURFER is provided, including identification of specific details of sources of computational bias

Nectin-2 expression in testicular cells is controlled via the functional cooperation between transcription factors of the Sp1, CREB, and AP-1 families

Nectin-2, a major protein component of the adherens junctions (AJs), is found between Sertoli cells and germ cells in the seminiferous epithelium. Recent studies have shown that the expression of nectin-2 gene in testis is crucial to maintain normal spermatogenesis since male knockout mice lacking nectin-2 gene are sterile and possess morphologically abnormal spermatozoa. However, the molecular mechanisms governing its basal transcription remain poorly understood. By the use of Sertoli and germ cell-lines (TM4 and GC-2spd(ts) cells, respectively) in transient transfection studies, we showed that the minimal mouse nectin-2 promoter was located between nucleotides -316 and -211 (relative to the translation start site). Two putative Sp1 motifs and one each of the CRE, AP1, and AP2 motifs were identified within this region. Mutational studies showed that these two Sp1 motifs cooperated synergistically with the CRE motif, but not the AP1 and AP2 motifs, to regulate nectin-2 gene transcription in both TM4 and GC-2spd(ts) cells. By EMSAs, we found that an AP-1 consensus sequence was able to inhibit DNA-protein complex formation with the CRE motif, suggesting an interaction between the AP-1 transcription factor (c-Jun) and CREB within the CRE motif. Overexpressions of CREB and c-Jun, but not c-Fos, also significantly increased the promoter activity, which suggests that CREB and c-Jun are the crucial transcription factors involved in regulating nectin-2 gene transcription. Chromatin immunoprecipitation assay has shown that, in vivo, CREB, c-Jun, and Sp1 family proteins are bound to the mouse nectin-2 promoter. Analysis of the staged tubules has confirmed that the cyclic expressions of CREB and nectin-2 coincide with the event of adherens junction restructuring between Sertoli cells and germ cells. The cross-talk between CREB, c-Jun, and Sp1 family protein is believed to be a major transcription machinery to drive nectin-2 expression in Sertoli cells.postprin

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts