This manual establishes the procedures for the collection of environmental samples and the performance of radiation surveys and other field measurements. Responsibilities are defined for those personnel directly involved in the collection of samples and the performance of field measurements.

This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.

The objective of this training is to: describe the responsibilities, resources, and goals of the Emergency Operations Center and be able to evaluate and interpret this information to best direct and allocate emergency, plant, and other resources to protect life and the Paducah Gaseous Diffusion Plant.

This report documents the updated six-degree-of-freedom reentry simulation TMAGRA6C used in the Aerospace Nuclear Safety Program, ANSP. The simulation provides for the inclusion of the effects of ablation on the aerodynamic stability and drag of reentry bodies, specifically the General Purpose Heat Source, GPHS. The existing six-degree-of-freedom reentry body simulations (TMAGRA6A and TMAGRA6B) used in the JHU/APL Nuclear Safety Program do not include aerodynamic effects resulting from geometric changes to the configuration due to ablation from reentry flights. A wind tunnel test was conducted in 1989 to obtain the effects of ablation on the hypersonic aerodynamics of the GPHS module. The analyzed data were used to form data sets which are included herein in tabular form. These are used as incremental aerodynamic inputs in the new TMAGRA6C six-degree-of-freedom reentry simulation. 20 refs., 13 figs., 2 tabs.

The 20,000 gls. Liquid Argon dewar stores up to 15,000 gls. of high purity (<1.0 ppm O{sub 2}, 0.999995) LAr for use in the Liquid Argon calorimeters of E740, the D0 collider detector, at elevation 707-feet. The dewar provides for the total detector volume of 11,000 gls and a 4,000 gls. storage inventory. The large gas volume ({ge}5,000 gls.) serves operational needs and guards against overfill concerns. The LAr dewar functions in two modes: (1) low pressure (16 psi relief) storage, and liquid and gas transfer operations to and from the low pressure (13 psi relief) detector cryostats, and (2) high pressure (65 psi relief) liquid transfer operations to and from a delivery trailer at elevation 743-feet. The storage function is intended to be long term and nonventing. The dewar is equipped with a 40 kW LN{sub 2} condenser that operates to maintain the pressure constant in the storage mode. This service exactly parallels the NeH{sub 2} and D{sub 2} storage dewar services provided at the 15-feet bubble chamber for its operation.

Survey report documenting the findings in the waters of Clear Fork Brazos River from August 22-25, 1988 and November 29- December 2, 1988.

The three boxes being reviewed all operate at a pressure of less than 15 psig. Since they are relieved at 13 psig, they fall outside the scopes of the ASME Pressure Vessel Code, Fermilab Engineering Standard SD-37B, and Chapter 5031 of the Fermilab Safety Manual, therefore a Pressure Vessel Engineering Note showing compliance with SD-37B is not required. In calculating the design stresses, only the largest of the three boxes, the signal board feed-thru box, was analyzed. This box had the largest spans and areas and would experience the largest pressure-related forces. The thinnest walls of each box were found to be in the top plates and they were also the side of the box which exposed the largest amount of area to internal pressure. The signal board feed-thru box top plate had at least twice the pressure area than either the instrumentation or hi-voltage boxes' top plates. This large disparity overshadows the slight difference in top plate thicknesses between the three boxes (0.56-inch vs. 0.25-inch and 0.3125-inch, respectively). Therefore, we felt the analysis of the larger signal board feedthru box would justify the design of the smaller instrumentation and hi-voltage boxes. Appended to the end of this engineering note …

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management.

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., …

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This document establishes the software design, nomenclature, and conventions for industrial Computed Tomography (CT) used in the Nondestructive Evaluation Section at Lawrence Livermore National Laboratory. It is mainly a users guide to the technical use of the CT computer codes, but also presents a proposed standard for describing CT experiments and reconstructions. Each part of this document specifies different aspects of the CT software organization. A set of tables at the end describes the CT parameters of interest in our project. 4 refs., 6 figs., 1 tab.

This report contains a safety analysis for the X17B2 beamline synchrotron medical research facility. Health hazards, risk assessment and building systems are discussed. Reference is made to transvenous coronary angiography. (LSP)

A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

Purpose of the Phase I effort was to demonstrate feasibility of the fuel cell/battery system for powering a small bus (under 30 ft or 9 m) on an urban bus route. A brassboard powerplant was specified, designed, fabricated, and tested to demonstrate feasibility in the laboratory. The proof-of-concept bus, with a powerplant scaled up from the brassboard, will be demonstrated under Phase II.

The Department of Energy (DOE) has prepared this report on the progress of site characterization activities at Yucca Mountain in southern Nevada. This report is the first of a series of reports that will hereafter be issued at intervals of approximately 6-months during site characterization. The DOE`s plans for site characterization are described in the Site Characterization Plan (SCP) for the Yucca Mountain site. The SCP has been reviewed and commented on by the NRC, the State of Nevada, the affected units of local government, other interested parties, and the public. More detailed information on plans for site characterization is being presented in study plans for the various site characterization activities. This progress report presents short summaries of the status of site characterization activities and cites technical reports and research products that provide more detailed information on the activities. The report provides highlights of work started during the reporting period, work in progress, and work completed and documented during the reporting period. In addition, the report is the vehicle for discussing major changes, if any, to the DOE`s site characterization program resulting from ongoing collection and evaluation of site information; the development of repository and waste-package designs; receipt of performance-assessment …

The purpose of this study is to fulfill the requirements of Public Law No. 101-46, approved June 30, 1989. The study describes and evaluates alternative methods for financing the future expansion of the Strategic petroleum Reserve (SPR), both to the current target level of 750 million barrels and to potential future levels of up to one billion barrels.

This report describes a material model driver which has recently been implemented in the DYNA3D code. The material model driver allows plotting of the constitutive response predicted by a material model under a given load path. This capability is particularly useful when fitting complex material models to experimental data. The plotting capability of the material model driver facilitates comparison of the simulated material stress-strain behavior with actual material test results. 1 ref., 6 figs., 4 tabs.

This radiologic characterization of tho two inactive uranium millsites at Rifle, Colorado, was conducted by Bendix Field Engineering Corporation (Bendix) for the US Department of Energy (DOE), Grand Junction Projects Office, in accord with a Statement of Work prepared by the DOE Uranium Mill Tailings Remedial Action (UMTRA) Project Technical Assistance Contractor, Jacobs Engineering Group, Inc. (Jacobs). The purpose of this project is to define the extent of radioactive contamination at the Rifle sites that exceeds US Environmental Protection Agency, (EPA) standards for UMTRA sites. The data presented in this report are required for characterization of the areas adjacent to the tailings piles and for the subsequent design of cleanup activities. An orientation visit to the study area was conducted on 31 July--1 August 1984, in conjunction with Jacobs, to determine the approximate extent of contaminated area surrounding tho piles. During that visit, survey control points were located and baselines were defined from which survey grids would later be established; drilling requirements were assessed; and radiologic and geochemical data were collected for use in planning the radiologic fieldwork. The information gained from this visit was used by Jacobs, with cooperation by Bendix, to determine the scope of work required for …

A review of the Savannah River Site (SRS) Probabilistic Risk Assessment (PRA) has been performed by a review committee organized by the US Department of Energy (DOE) and its contractor, EG&G Idaho, Inc. The High-Level Peer Review Committee (referred to as ``the Committee`` in this report) members are identified in Section 2. The main purpose of the review has been to provide assurance that the SRS PRA is responsive to safety issues associated with the restart and continued operation of the Savannah River reactors. The Committee members are all experienced practitioners of PRA, and several of the members have been deeply involved In a concurrent, detailed review of the SRS PRA. Source material and expertise available to the Committee included the SRS PRA document itself issued August 31. 1989, and Interaction with key PRA and plant experts at both the Savannah River Site and the Los Alamos National Laboratory (LANL), who had performed an independent PRA evaluation of the SRS K-reactor. The cooperation and support received from those connected with the review were outstanding.

The pilot-scale ceramic melter test, was conducted to determine the vitrification processing characteristics of simulated Hanford Waste Vitrification Plant process slurries and the integrated performance of the melter off-gas treatment system. Simulated melter feed was prepared and processed to produce glass. The vitrification system, achieved an on-stream efficiency of greater than 98%. The melter off-gas treatment system included a film cooler, submerged bed scrubber, demister, high-efficiency mist eliminator, preheater, and high-efficiency particulate air filter (HEPA). Evaluation of the off-gas system included the generation, nature, and capture efficiency of gross particulate, semivolatile, and noncondensible melter products. 17 refs., 48 figs., 61 tabs.

Risk Management is the optimization of safety programs. This requires a formal systems approach to hazards identification, risk quantification, and resource allocation/risk acceptance as opposed to case-by-case decisions. The Management Oversight and Risk Tree (MORT) has gained wide acceptance as a comprehensive formal systems approach covering all aspects of risk management. It (MORT) is a comprehensive analytical procedure that provides a disciplined method for determining the causes and contributing factors of major accidents. Alternatively, it serves as a tool to evaluate the quality of an existing safety system. While similar in many respects to fault tree analysis, MORT is more generalized and presents over 1500 specific elements of an ideal ''universal'' management program for optimizing occupational safety.

This document contains the findings and associated root causes identified during the Tiger Team Assessment of the Department of Energy's (DOE) Pantex Plant in Amarillo, Texas. This assessment was conducted by the Department's Office of Environment, Safety and Health between October 2 and 31, 1989. The scope of the assessment of the Pantex Plant covered all areas of environment, safety and health (ES H) activities, including compliance with federal, state, and local regulations, requirements, permits, agreements, orders and consent decrees, and DOE ES H Orders. The assessment also included an evaluation of the adequacy of DOE and site contractor ES H management programs. The draft findings were submitted to the Office of Defense Programs, the Albuquerque Operations Office, the Amarillo Area Office, and regulatory agencies at the conclusion of the on-site assessment activities for review and comment on technical accuracy. Final modifications and any other appropriate changes have been incorporated in the final report. The Tiger Team Assessment of the Pantex Plant is part of the larger Tiger Team Assessment program which will encompass over 100 DOE operating facilities. The assessment program is part of a 10-point initiative announced by Secretary of Energy James D. Watkins on June 27, 1989, …

In this note, single and multi-bunch collective instabilities are considered for PEP operating a 9 {times} 9 bunch configuration. The lattice is based on a vertically separated beam pretzel' design which allows for collisions at the TPC (IR2) only. Threshold current levels and linear instability growth rates are calculated with the storage ring design code {prime}ZAP{prime}. Single bunch instabilities should not be a problem for total circulating currents of 100mA (18 bunches, 5.6mA/bunch). Coupled-bunch growth rate calculations are based on a line broadening technique for the higher-order cavity modes. In the longitudinal case, feedback will be required. For the transverse coupled bunch instabilities, growth rates are about 5 times less. 14 refs., 5 figs.