Comparison of elasticity and shell-theory solutions for finite circular cylindrical shells

A finite circular cylindrical shell subjected to a band of uniform pressure on its outer rim was investigated, using three-dimensional elasticity theory and the classical shell theories of Timoshenko (or Donnell) and Flügge. Detailed comparison of the resulting stresses and displacements was carried out for shells with ratios of inner to outer shell radii equal to 0.80, 0.85, 0.90 and 0.93 and for ratios of outer shell diameter to length of the shell equal to 0.5, 1 and 2. The ratio of band width to length of the shell was 0.2 and Poisson's ratio used was equal to 0.3. An Elliot 803 digital computer was used for numerical computations.

Plastic response of orthotropic spherical shells under blast loading

The plastic response of a segment of a simply supported orthotropic spherical shell under a uniform blast loading applied on the convex surface of the shell is presented. The blast is assumed to impart a uniform velocity to the shell surface initially. The material of the shell is orthotropic obeying a modified Tresca yield hypersurface conditions and the associated flow rules. The deformation of the shell is determined during all phases of its motion by considering the motion of plastic hinges in different regimes of flow. Numerical results presented include the permanent deformed configuration of the shell and the total time of shell response for different degrees of orthotropy. Conclusions regarding the plastic behaviour of spherical shells with circumferential and meridional stiffening under uniform blast load are presented.

Thermal stresses in a spherical shell with a conical nozzle

A thermal stress problem of a spherical shell with a conical nozzle is solved using a continuum approach. The thermal loading consists of a steady temperature which is uniform on the inner and outer surfaces of the shell and the conical nozzle but may vary linearly across the thickness. The thermal stress problem is converted to an equivalent boundary value problem and boundary conditions are specified at the junction of the spherical shell and conical nozzle. The stresses are obtained for a uniform increase in temperature and for a linear variation of temperature across the thickness of the shell, and are presented in graphical form for ready use.

Advances in the analysis and design of concrete structures, metal containments and liner plate for extreme loads

The material presented in this paper summarizes the progress that has been made in the analysis, design, and testing of concrete structures. The material is summarized in the following documents: 1. Part I - Containment Design Criteria and Loading Combinations - J.D. Stevenson (Stevenson and Associates, Cleveland, Ohio, USA) 2. Part II - Reinforced and Prestressed Concrete Behavior - J. Eibl and M. Curbach (Karlsruhe University, Karlsruhe, Germany) 3. Part III - Concrete Containment Analysis, Design and Related Testing - T.E. Johnson and M.A. Daye (Bechtel Power Corporation, Gaithersburg, Maryland USA) 4. Part IV - Impact and Impulse Loading and Response Prediction - J.D. Riera (School of Engineering - UFRGS, Porto Alegre, RS, Brazil) 5. Part V - Metal Containments and Liner Plate Systems - N.J. Krutzik (Siemens AG, Offenbach Am Main, Germany) 6. Part VI - Prestressed Reactor Vessel Design, Testing and Analysis - J. Nemet (Austrian Research Center, Seibersdorf, Austria) and K.T.S. Iyengar (Indian Institute of Science, Bangalore, India).

Thermomechanical generalization of pin joints in finite plates

The understanding of thermoelastic behaviour of joints is significant in order to ensure the integrity of large and complex structures exposed to a thermal environment, particularly in fields such as aerospace and nuclear engineering. Thermomechanical generalization of partial contact behaviour of a pin joint under combined in-plane mechanical loading and on-axis unidirectional heat flow has already been established by the authors for the analytically simpler domains of large plates. This paper successfully extends the on-going investigation to a single pin in a finite rectangular isotropic plate as a two-dimensional abstraction from a practical situation of a multipin fastener joint. The finite element method is used to analyse the joint problem under on-axis thermomechanical loading and unified load-contact relationships are established for a class of loading conditions.

Effective thermal conductivity of a heat generating rod bundle dissipating heat by natural convection and radiation

A numerical study of conjugate natural convection and surface radiation in a horizontal hexagonal sheath housing 19 solid heat generating rods with cladding and argon as the fill gas, is performed. The natural convection in the sheath is driven by the volumetric heat generation in the solid rods. The problem is solved using the FLUENT CFD code. A correlation is obtained to predict the maximum temperature in the rod bundle for different pitch-to-diameter ratios and heat generating rates. The effective thermal conductivity is related to the heat generation rate, maximum temperature and the sheath temperature. Results are presented for the dimensionless maximum temperature, Rayleigh number and the contribution of radiation with changing emissivity, total wattage and the pitch-to-diameter ratio. In the simulation of a larger system that contains a rod bundle, the effective thermal conductivity facilitates simplified modelling of the rod bundle by treating it as a solid of effective thermal conductivity. The parametric studies revealed that the contribution of radiation can be 38-65% of the total heat generation, for the parameter ranges chosen. Data for critical Rayleigh number above which natural convection comes into effect is also presented. (C) 2011 Elsevier B.V. All rights reserved.

Estimation of design basis earthquake using region-specific M-max, for the NPP site at Kalpakkam, Tamil Nadu, India

The objective of the paper is to estimate Safe Shutdown Earthquake (SSE) and Operating/Design Basis Earthquake (OBE/DBE) for the Nuclear Power Plant (NPP) site located at Kalpakkam, Tamil Nadu, India. The NPP is located at 12.558 degrees N, 80.175 degrees E and a 500 km circular area around NPP site is considered as `seismic study area' based on past regional earthquake damage distribution. The geology, seismicity and seismotectonics of the study area are studied and the seismotectonic map is prepared showing the seismic sources and the past earthquakes. Earthquake data gathered from many literatures are homogenized and declustered to form a complete earthquake catalogue for the seismic study area. The conventional maximum magnitude of each source is estimated considering the maximum observed magnitude (M-max(obs)) and/or the addition of 0.3 to 0.5 to M-max(obs). In this study maximum earthquake magnitude has been estimated by establishing a region's rupture character based on source length and associated M-max(obs). A final source-specific M-max is selected from the three M-max values by following the logical criteria. To estimate hazard at the NPP site, ten Ground-Motion Prediction Equations (GMPEs) valid for the study area are considered. These GMPEs are ranked based on Log-Likelihood (LLH) values. Top five GMPEs are considered to estimate the peak ground acceleration (PGA) for the site. Maximum PGA is obtained from three faults and named as vulnerable sources to decide the magnitudes of OBE and SSE. The average and normalized site specific response spectrum is prepared considering three vulnerable sources and further used to establish site-specific design spectrum at NPP site.

Hydro-mechanical characterization of Barmer 1 bentonite from Rajasthan, India

The conceptual model for deep geological disposal of high level nuclear waste (HLW) is based on multiple barrier system consisting of natural and engineered barriers. Buffer/backfill material is regarded as the most important engineered barrier in HLW repositories. Due to large swelling ability, cation adsorption capacity, and low permeability bentonite is considered as suitable buffer material in HLW repositories. Japan has identified Kunigel VI bentonite, South Korea - Kyungju bentonite, China - GMZ bentonite, Belgium - FoCa clay, Sweden - MX-80 bentonite, Spain - FEBEX bentonite and Canada - Avonseal bentonite as candidate bentonite buffer for deep geological repository program. An earlier study on Indian bentonites by one of the authors suggested that bentonite from Barmer district of Rajasthan (termed Barmer 1 bentonite), India is suited for use as buffer material in deep geological repositories. However, the hydro-mechanical properties of the Barmer 1 bentonite are unavailable. This paper characterizes Barmer 1 bentonite for hydro-mechanical properties, such as, swell pressure, saturated permeability, soil water characteristic curve (SWCC) and unconfined compression strength at different dry densities. The properties of Barmer 1 bentonite were compared with bentonite buffers reported in literature and equations for designing swell pressure and saturated permeability coefficient of bentonite buffers were arrived at. (C) 2013 Elsevier B.V. All rights reserved.

A method for VVER-1000 fuel rearrangement optimization taking into account both fuel cladding durability and burnup

A method for VVER-1000 fuel rearrangement optimization that takes into account both cladding durability and fuel burnup and which is suitable for any regime of normal reactor operation has been established. The main stages involved in solving the problem of fuel rearrangement optimization are discussed in detail. Using the proposed fuel rearrangement efficiency criterion, a simple example VVER-1000 fuel rearrangement optimization problem is solved under deterministic and uncertain conditions. It is shown that the deterministic and robust (in the face of uncertainty) solutions of the rearrangement optimization problem are similar in principle, but the robust solution is, as might be anticipated, more conservative. © 2013 Elsevier B.V.

Core design options for high conversion BWRs operating in Th- 233U fuel cycle

Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th-233U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone "sandwiched" between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design. © 2013 Elsevier B.V. All rights reserved.

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results. Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al.; 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module. Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions. © 2011 Published by Elsevier B.V.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX. © 2009 Elsevier B.V. All rights reserved.

Decay heat in fast reactors with transuranic fuels

In this work, we performed an evaluation of decay heat power of advanced, fast spectrum, lead and molten salt-cooled reactors, with flexible conversion ratio. The decay heat power was calculated using the BGCore computer code, which explicitly tracks over 1700 isotopes in the fuel throughout its burnup and subsequent decay. In the first stage, the capability of the BGCore code to accurately predict the decay heat power was verified by performing a benchmark calculation for a typical UO2 fuel in a Pressurized Water Reactor environment against the (ANSI/ANS-5.1-2005, "Decay Heat Power in Light Water Reactors," American National Standard) standard. Very good agreement (within 5%) between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power for fast reactors with different coolants and conversion ratios, for which no standard procedure is currently available. Notable differences were observed for the decay power of the advanced reactor as compared with the conventional UO2 LWR. The importance of the observed differences was demonstrated by performing a simulation of a Station Blackout transient with the RELAP5 computer code for a lead-cooled fast reactor. The simulation was performed twice: using the code-default ANS-79 decay heat curve and using the curve calculated specifically for the studied core by BGCore code. The differences in the decay heat power resulted in failure to meet maximum cladding temperature limit criteria by ∼100 °C in the latter case, while in the transient simulation with the ANS-79 decay heat curve, all safety limits were satisfied. The results of this study show that the design of new reactor safety systems must be based on decay power curves specific to each individual case in order to assure the desired performance of these systems. © 2009 Elsevier B.V. All rights reserved.