WorldCat Identities

United States Office of Nuclear Energy, Science, and Technology

Overview
Works: 1,876 works in 1,925 publications in 1 language and 8,007 library holdings
Genres: Periodicals 
Roles: Sponsor, Researcher
Classifications: TK9202, 621.4830973
Publication Timeline
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Most widely held works by United States
Annual report by Nuclear Energy Research Initiative (U.S.)( )

in English and held by 214 WorldCat member libraries worldwide

Annual report by International Nuclear Energy Research Initiative (U.S.)( )

in English and held by 198 WorldCat member libraries worldwide

University currents( )

in English and held by 141 WorldCat member libraries worldwide

Nuclear reactors built, being built, or planned in the United States as of( )

in English and held by 116 WorldCat member libraries worldwide

Draft environmental impact statement for the proposed consolidation of nuclear operations related to production of radioisotope power systems by United States( Book )

1 edition published in 2005 in English and held by 12 WorldCat member libraries worldwide

Final environmental impact statement for the treatment and management of sodium-bonded spent nuclear fuel : summary( Book )

1 edition published in 2000 in English and held by 8 WorldCat member libraries worldwide

Advanced Fuel Cycle Cost Basis( )

3 editions published between 2007 and 2009 in English and held by 0 WorldCat member libraries worldwide

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules--24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste
PEBBLES Operation and Theory Manual( )

2 editions published between 2010 and 2011 in English and held by 0 WorldCat member libraries worldwide

The PEBBLES manual describes the PEBBLES code. The PEBBLES code is a computer program designed to simulation the motion, packing and vibration of spheres that undergo various mechanical forces including gravitation, Hooke's law force and various friction forces. The frictional forces include true static friction that allows non-zero angles of repose. Each pebble is individually simulated using the distinct element method
EVALUATION OF THE START-UP CORE PHYSICS TESTS AT JAPAN'S HIGH TEMPERATURE ENGINEERING TEST REACTOR (FULLY-LOADED CORE)( )

2 editions published between 2009 and 2010 in English and held by 0 WorldCat member libraries worldwide

Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume VIII. Advanced concepts( )

2 editions published between 1979 and 1980 in English and held by 0 WorldCat member libraries worldwide

The six advanced concepts for nuclear power systems that were selected for evaluation are: the fast mixed-spectrum reactor; the denatured molten-salt reactor; the mixed-flow gaseous-core reactor; the linear-accelerator fuel-regenerator reactor; the ternary metal-fueled electronuclear fuel-producer reactor; and the tokamak fusion-fission hybrid reactor. The design assessment was performed by identifying needs in six specific areas: conceptual plant design; reactor-physics considerations; fuel cycle alternatives; mechanical and thermal-hydraulic considerations; selection, development, and availability of materials; and engineering and operability. While none of the six concepts appears to be a credible commercial alternative to the liquid-metal fast-breeder within the Nonproliferation Alternative Systems Assessment Program horizon of 2025, there are a number of reasons for continued interest in the fast mixed-spectrum reactor: it is a once-through cycle fast reactor with proliferation risk characteristics similar to those of the light-water reactor; only about one-third as much uranium is required as for the once-through light-water reactor; the system will benefit directly from fast-breeder development programs; and, finally, the research and development required to develop the high-burnup metal fuel could benefit the on-going liquid-metal fast-breeder reactor program. Accordingly, a limited research and development effort on the high-burnup fuel seems justified, at present
Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis( )

3 editions published between 2009 and 2011 in English and held by 0 WorldCat member libraries worldwide

This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy's mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site
Power Burst Facility( )

3 editions published in 1990 in English and held by 0 WorldCat member libraries worldwide

This report discusses the monthly progress of the Power Burst Facility/Boron Neutron Capture Therapy (PBF/BNLT) program for cancer treatment. Highlights of the PBF/BNCT Program during July 1990 include progress within the areas of: Gross boron analysis in tissue, blood, and urine; noninvasive boron quantitative determination; analytical radiation transport and interaction modeling for BNCT; large animal model studies; neutron source and facility preparation; administration and common support and PBF operations
A new balance-of-plant model for the SASSYS-1 LMR systems analysis code( )

2 editions published in 1989 in English and held by 0 WorldCat member libraries worldwide

Models of power plant heat transfer components and rotating machinery have been added to the balance-of-plant model in the SASSYS-1 liquid metal reactor systems analysis code. This work is part of a continuing effort in plant network simulation based on the general mathematical models developed. The models described in this paper extend the scope of the balance-of-plant model to handle non-adiabatic conditions along flow paths. While the mass and momentum equations remain the same, the energy equation now contains a heat source term due to energy transfer across the flow boundary or to work done through a shaft. The heat source term is treated fully explicitly. In addition, the equation of state is rewritten in terms of the quality and separate parameters for each phase. The models are simple enough to run quickly, yet include sufficient detail of dominant plant component characteristics to provide accurate results. 5 refs., 16 figs., 2 tabs
CFD Analysis of Core Bypass Phenomena( )

2 editions published between 2009 and 2010 in English and held by 0 WorldCat member libraries worldwide

The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary
A method for pressure-pulse suppression in fluid-filled piping( )

2 editions published between 1989 and 1990 in English and held by 0 WorldCat member libraries worldwide

A simple, nondestructive method to suppress pressure pulses in fluid-filled piping was proposed and theoretically analyzed earlier. In this paper, the proposed method is verified experimentally. The results of experiments performed for the range of parameters of practical importance indicated that the attenuation of pressure pulses was in accordance with the theoretical predictions. This paper describes the experimental setup and the test models of the proposed pulse suppression devices and discusses the experimental results. In particular, the measured attenuation factors are presented and compared with the theoretical predictions. 8 ref., 17 fig., 2 tab
Next Generation Nuclear Plant System Requirements Manual( )

3 editions published between 2007 and 2009 in English and held by 0 WorldCat member libraries worldwide

Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program( )

7 editions published in 1990 in English and held by 0 WorldCat member libraries worldwide

The reactor delta T'', the difference between the average core inlet and outlet temperatures, for the liquid-sodium-cooled Experimental Breeder Reactor 2 is empirically synthesized in real time from, a multitude of examples of past reactor operation. The real-time empirical synthesis is based on reactor operation. The real-time empirical synthesis is based on system state analysis (SSA) technology embodied in software on the EBR 2 data acquisition computer. Before the real-time system is put into operation, a selection of reactor plant measurements is made which is predictable over long periods encompassing plant shutdowns, core reconfigurations, core load changes, and plant startups. A serial data link to a personal computer containing SSA software allows the rapid verification of the predictability of these plant measurements via graphical means. After the selection is made, the real-time synthesis provides a fault-tolerant estimate of the reactor delta T accurate to +/-1%. 5 refs., 7 figs
Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors( )

2 editions published in 1990 in English and held by 0 WorldCat member libraries worldwide

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab
 
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Alternative Names

controlled identityUnited States. Department of Energy

United States. Department of Energy. Office of Nuclear Energy, Science, and Technology

Languages
English (52)