The J-PARC Neutrino Experiment, the construction of which starts in JFY 2004, will use a superconducting magnet system for its primary proton beam line. The system, which bends the 50 GeV 0.75 MW proton beam by about 80 degrees, consists of 28 superconducting combined function magnets. The magnets utilize single layer left/right asymmetric coils that generate a dipole field of 2.6 T and a quadrupole field of 18.6 T/m with the operation current of about 7.35 kA. The system also contains a few conduction cooled superconducting corrector magnets that serve as vertical and horizontal steering magnets. All the magnets are designed to provide a physical beam aperture of 130 mm in order to achieve a large beam acceptance. Extensive care is also required to achieve safe operation with the high power proton beam. The paper summarizes the system design as well as some safety analysis results.

Extremely high radiation, levels with accumulated doses comparable to those in nuclear reactors than in accelerators, and very high heat loads ({approx}15 kw) make the quadrupole magnets in the fragment separator one of the most challenging elements of the proposed Rare Isotope Accelerator (RIA). Removing large heat loads, protecting the superconducting coils against quenching, the long term survivability of magnet components, and in particular, insulation that can retain its functionality in such a harsh environment, are the major challenges associated with such magnets. A magnet design based on commercially available high temperature superconductor (HTS) and stainless steel tape insulation has been developed. HTS will efficiently remove these large heat loads and stainless steel can tolerate these large radiation doses. Construction of a model magnet has been started with several coils already built and tested. This paper presents the basic magnet design, results of the coil tests, the status and the future plans. In addition, preliminary results of radiation calculations are also presented.

Semi-weekly newspaper from Livingston, Texas that includes local, state and national news along with advertising.

The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

This presentation discusses the changes in nuclear engineering education.

The Absolute Neutron Activation Analysis (ANAA) technique was used to determine element concentrations of Cl and Na in blood of healthy group (male and female blood donators), select from Blood Banks at Sao Paulo city, to provide information which can help in diagnosis of patients. This study permitted to perform a discussion about the advantages and limitations of using this nuclear methodology in hematological examinations.

This paper discusses the implementation of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The Joint Convention: establishes a commitment with respect to safe management of spent nuclear fuel and radioactive waste; requires the Parties to ''take appropriate steps'' to ensure the safety of their spent fuel and waste management activities, but does not delineate standards the Parties must meet; and seeks to attain, through its Contracting Parties, a higher level of safety with respect to management of their spent nuclear fuel, disused sealed sources, and radioactive waste.

The IRIS (International Reactor Innovative and Secure) reactor design is being developed by an international consortium of 21 organizations from ten countries, including three members from Brazil and one from Mexico. This reflects the interest that Latin America has for a project which addresses the energy needs of the region. Presented here are some of the most recent developments in the IRIS project. The project's highest priority is the current pre-application licensing with the US NRC, which has required an investigation of the major accident sequences and a preliminary probabilistic risk assessment (PRA). The results of the accident analyses confirmed the outstanding inherent safety of the IRIS configuration and the PRA analyses indicated a core damage frequency due to internal events of the order of 2E-8. This not only highlights the enhanced safety characteristic of IRIS which should enhance its public acceptance, but it has also prompted IRIS to consider the possibility of being licensed without the need for off-site emergency response planning which would have a very positive economic implication. The modular IRIS, with each module rated at {approx} 335 MWe, is of course an ideal size for developing countries as it allows to easily introduce a moderate amount …

The smelting process associated with the advanced spent fuel conditioning process (ACP) of Korea Atomic Energy Research Institute was studied by using surrogate materials. Considering the vaporization behaviors of input materials, the operation procedure of smelting was set up as (1) removal of residual salts, (2) melting of metal powder, and (3) removal of dross from a metal ingot. The behaviors of porous MgO crucible during smelting were tested and the chemical stability of MgO in the salt-being atmosphere was confirmed.

The Isotope Production Facility (IPF) at Los Alamos is operated on the authorization basis of a radiological facility with an inventory limit of a Category 3 Nuclear Facility. For the commissioning of IPF, a ''dummy'' target stack containing Zn, Nb and Al disks, and a ''prototype'' stack were irradiated with a proton beam. The ''prototype'' stack contained two pressed RbCl disks, encapsulated in stainless steel, and a Ga metal target. Typical ''prototype'' stack beam parameters were 88.9 {micro}A, 101.3 h. Operation procedures require the projection of all generated radionuclide activities. This is mandatory in order to determine both maximum beam current and maximum beam exposure time. The Monte Carlo code MCNPX and the burn-up code CINDER90 were used to determine maximum beam parameters prior to irradiation. After irradiation, activity estimates were calculated assuming actual average beam parameters. They were entered into an online inventory database, and were later, after chemical separation and radioactive assays, replaced by experimental values. A comparison of ''prototype'' stack experimental yield data to Monte Carlo calculation results showed that the computer codes provide realistic, conservative estimates.

Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties.

The Anadenathera macrocarpa, Schinus molle, Hymenaea courbaril, Cariniana legalis, Solidago microglossa and Stryphnodendron barbatiman, were collected ''in natura'' samples (leaves, flowers, barks and seeds) from different commercial suppliers. The pharmaco-active compounds in ethanolic extracts had been made by the Mato Grosso Federal University (UFMT). The energy-dispersive x-ray fluorescence (ED-XRF) spectrometry was used for the elemental analysis in different parts of the plants and respective ethanolic extracts. The Ca, Cl, Cu, Fe, K, Mg, Mn, Na, Ni, P, Rb, S, Sr and Zn concentrations were determined by the fundamental parameters method. Some specimens showed a similar inorganic profile for ''in natura'' and ethanolic extract samples and some ones showed a distinct inorganic profile. For example, the Anadenathera macrocarpa showed a similar concentration in Mg, P, Cu, Zn and Rb elements in ''in natura'' and ethanolic extract samples; however very different concentration in Na, S, Cl, K , Ca, Mn, Fe and Sr was observed in distinctive samples. The Solidago microglossa showed the K, Ca, Cl, S, Mg, P and Fe elements as major constituents in both samples, suggesting that the extraction process did not affect in a considerable way the ''in natura'' inorganic composition. The elemental composition of the different parts …

The recovery of Cs and Sr from acidic solutions by solvent extraction has been investigated. The goal of this project was to develop an extraction process to remove Cs and Sr from high-level waste in an effort to reduce the heat loading in storage. Solvents for the extraction of Cs and Sr separately have been used on both caustic and acidic spent nuclear fuel waste in the past. The objective of this research was to find a suitable solvent for the extraction of both Cs and Sr simultaneously from acidic nitrate media.

This presentation discusses external costs associated with electricity generation options in Brazil.

This presentation discusses applications of radioisotopes in modern life.

During the last fifty years remarkable results are achieved in the application of nuclear technology for the production of electricity. Looking ahead to the next fifty years it is clear that the demand for energy will grow considerably and also the requirements for the way the energy will be supplied. Within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), the future of the energy demand and supply was explored and several scenario's identified. A leading requirement for energy supply is coming up and will play a crucial role: sustainability of the way the energy supply will be realized. Fulfilling the growing need for energy in developing countries is as well an important issue. Based on these scenario's for the next fifty years, an inventory of requirements for the future of nuclear energy systems has been collected as well a methodology developed by INPRO to assess innovative nuclear systems and fuel cycles. On the base of this assessment, the need for innovations and breakthroughs in existing technology can be defined. To facilitate the deployment of innovative nuclear systems also the infrastructure, technical as well as institutional has to be adjusted to the anticipated changes in the world such …

This paper summarizes the current comprehensive comparison of four major fuel cycle strategies: once-through, thermal recycle, thermal+fast recycle, fast recycle. It then proceeds to summarize comparison of the major technology options for the key elements of the fuel cycle that can implement each of the four strategies - separation processing, transmutation reactors, and fuels.

This presentation discusses the Integrated Management Program for Radioactive Realed Sources (IMPRSS) in Egypt.

The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas.

This presentation discusses the sharing of nuclear safety information through distributed database systems.

The aimed of this work consists of evaluating the influence of the dose secondary components (thermal neutrons dose, epithermal neutrons dose, fast neutrons dose and photon dose) in treatment planning with BNCT. MCNP4B Code was used to calculate RBE-Gy doses through the irradiation of the modified Snyder head phantom. A reduction of the therapeutical gain of monoenergetic neutron beans was observed in non invasive treatments, provoked for the predominance of the fast neutron dose component in the skin, showing that the secondary components of dose can to contribute more for to raise the healthy-tissue dose of that in the tumor, reducing the treatment efficiency.

The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; …

This paper discusses ten years of public acceptance experience in transports for Europe and coastal states between France and Japan and examples of Central and South America.

The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository. There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected.