Seismic Characterisation of Soils with SPT: Comparison of Calculated vs Values and Measured vs Values

Italian seismic regulatory guidance (O.P.C.M. n. 3274 of 20 March 2003, O.P.C.M. n. 3519 of 28 April 2006, implemented within the “Norme Tecniche per le Costruzioni” – D.M. of 14 January 2008) which conforms to European guidance (Eurocode 8), advises a seismic classification of foundation soils through the Vs30 value or alternatively the Nspt value. Several correlations between the Nspt parameter and the Vs parameter are known in literature (Otha & Goto, 1978; Imai & Tonouchi, 1982; Ohsaki & Iwasaki, 1977; Seed, 1983). Each of these correlations can only be applied in certain geological-technical contexts. The validity of these correlations should be discussed considering that geophysical investigations are undertaken under elastic deformation conditions where induced stress causes reversible deformations. Instead, SPT investigations cause rupture of the investigated material and they cannot be conducted within all types of deposits. This note presents the results obtained by comparison of Vs values from SPT investigations and the Vs values from down-hole investigations measured on the same vertical and at the same depth. The study is based on a great quantity of data (153 pairs of Nspt-Vs values) which were acquired from different geological-technical contexts. The data were acquired within the VEL project of the Tuscany Region (Central Italy) aimed at assessing local effects in areas where different seismic hazards exist. Limits and applicability of these empirical correlations are discussed

Argonne National Laboratory Reports

The SYNBURN computer program for fast reactors will calculate all the neutronics necessary to completely characterize the equilibrium cycle as well as the startup to equilibrium cycles. The program's run time is very short and this makes the program suitable for survey of parametric studies. It can search on the cycle time for a specified burnup, for the shim control necessary for criticality as well as feed enrichments and the enrichment ratio among core zones. SYNBURN synthesizes in a very simple fashion the one-dimensional fluxes in radial and axial geometry to achieve an approximate two-dimensional solution which agrees very well with the exact two-dimensional solution when measuring regional integrated quantities

Argonne National Laboratory Reports

Studies of unprotected loss-of-flow accidents in the CRBR for various rates of flow coast-down and with various options in the SAS 3A code did not lead to conditions for a violent disassembly. Maximum fuel temperatures using the SLUMPY module for disassembly were in the range 4000-4500 deg C. An approximate treatment of the LOF-driven TOP accident, not properly modeled by SAS 3A, indicates the possibility of some increase in accident severity. The effect of fission gas in dispersing fuel was not taken into account in these calculations. Parameter variations included the presence or absence of axial fuel expansion and of clad motion and use of the moving coolant film model versus the static film model. Study of severe pipe rupture accidents with scram indicated that pin power density and fuel-clad conductance were important parameters in determining what coolant flow rate was needed to prevent boiling after the rupture. It appears that for the CRBR when engineering hot channel factors are considered, this fraction would have to exceed 25 percent

Measurement of seismic local effects and 1D numerical modeling at selected sites affected by the 2009 seismic sequence of L’Aquila

Strong local effects occurred during the seismic sequence that affected the area of L’Aquila (Central Italy) in 2009, which culminated with a main shock of magnitude (Mw) 6.3 on the 6 April. Bearing in mind the magnitude, the vicinity of a regional capital to the epicentre, the presence of several historic centers and villages, which caused significant damage (IMCS max=9.5) and approximately 300 victims, among the main characteristics of this earthquake, was the variability of the recorded damages within the scale of 100 m in spite of the near field conditions and a distinct source mechanism. These local effects were measured with an array of geophones at selected sites near the city of L’Aquila. In one of the sites, located at Onna village which experienced maximum damage levels, the 1D numerical modeling allowed verifying the influence that the frequencey of the input motion has in the determination of superficial seismic effects. The numerical modeling, which was undertaken on the basis of the results from geomechanical, seismic and geoelectrical surveys, also allowed verifying that underestimation of local seismic amplification can be introduced by the current regulatory framework (Italian seismic regulatory guidance “Norme Tecniche per le Costruzioni” – D.M. of 14 January 2008) in contexts similar to that of L’Aquila

Organization and management of the plant safety evaluation of the VVER-440/230 units at Novovoronezh.

As part of the Soviet-Designed Reactor Safety (SDRS) element of the International Nuclear Safety Program (INSP), the US Department of Energy (US DOE) is funding a plant safety evaluation (PSE) project for the Novovoronezh Nuclear Power Plant (NvNPP). The Novovoronezh PSE Project is a multi-faceted project with participants from sixteen different international organizations from five different countries scattered across eleven time zones. The purpose of this project is to provide a thorough Probabilistic Risk Analysis (PRA) and Deterministic Safety Analysis (DSA) for Units 3 and 4 of the NvNPP. In addition, this project provides assistance to the operation organizations in meeting their international commitments in support of safety upgrades, and their regulatory requirements for the conduct of safety analyses. Managing this project is a complex process requiring numerous management tools, constant monitoring, and effective communication skills. Employing management tools to resolve unanticipated problems one of the keys to project success. The overall scope, programmatic context, objectives, project interactions, communications, practical hindrances, and lessons learned from the challenging performance of the PSE project are summarized in this paper

Analysis of fluid-structure interaction and structural response of Chernobyl-4 reactor

On April 26, 1986, an accident occurred at the Chernobyl-4 Nuclear Power Plant in the Soviet Union. A post accident meeting was held in Vienna during the week of August 25, 1986. In mid-July 1986, the DOE formed a team to analyze the accident, including experts from the national laboratories such as Argonne National Laboratory, Brookhaven National Laboratory, Oak Ridge National Laboratory, and Pacific Northwest Laboratory. The goal was to assess the information's plausibility, provided analytical support to the US delegation during the post-accident review meeting and obtain a technical understanding of the accident. Detailed analyses of the team work are given in Ref. 1 (DOE, 1986). The accident at Chernobyl-4 occurred during the running of a test to determine a turbogenerator's ability to provide in-house emergency power after shutting off its steam supply. The accident was the result of a large, destructive power excursion. The major design related factor in the accident was the large positive void coefficient of reactivity. This feature, not present in the US reactors, means that an increase in power is likely to lead to an increase in reactivity which will further increase power, and finally result in the destructive accident. 5 refs., 11 figs

Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of {minus}3$to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

Comparative Analysis of a Hypothetical Loss-of-Flow Accident in an Irradiated LMFBR Core Using Different Computer Models for a Common Benchmark Problem.

THIS REPORT SUMMARIZES THE RESULTS OF AN INTERNATIONAL EXERCISE TO COMPARE WHOLE-CORE ACCIDENT CALCULATIONS OF THE INITIATION PHASE OF AN UNPROTECTED LOF ACCIDENT IN A LARGE IRRADIATED LMFBR.NA-NOT AVAILABL