The role of the PIRT process in identifying code improvements and executing code development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, G.E.; Boyack, B.E.

1997-07-01

In September 1988, the USNRC issued a revised ECCS rule for light water reactors that allows, as an option, the use of best estimate (BE) plus uncertainty methods in safety analysis. The key feature of this licensing option relates to quantification of the uncertainty in the determination that an NPP has a {open_quotes}low{close_quotes} probability of violating the safety criteria specified in 10 CFR 50. To support the 1988 licensing revision, the USNRC and its contractors developed the CSAU evaluation methodology to demonstrate the feasibility of the BE plus uncertainty approach. The PIRT process, Step 3 in the CSAU methodology, wasmore » originally formulated to support the BE plus uncertainty licensing option as executed in the CSAU approach to safety analysis. Subsequent work has shown the PIRT process to be a much more powerful tool than conceived in its original form. Through further development and application, the PIRT process has shown itself to be a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. Used early in research directed toward these objectives, PIRT results also provide the technical basis and cost effective organization for new experimental programs needed to improve the safety analysis codes for new applications. The primary purpose of this paper is to describe the generic PIRT process, including typical and common illustrations from prior applications. The secondary objective is to provide guidance to future applications of the process to help them focus, in a graded approach, on systems, components, processes and phenomena that have been common in several prior applications.« less

Progress of IRSN R&D on ITER Safety Assessment

NASA Astrophysics Data System (ADS)

Van Dorsselaere, J. P.; Perrault, D.; Barrachin, M.; Bentaib, A.; Gensdarmes, F.; Haeck, W.; Pouvreau, S.; Salat, E.; Seropian, C.; Vendel, J.

2012-08-01

The French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN), in support to the French "Autorité de Sûreté Nucléaire", is analysing the safety of ITER fusion installation on the basis of the ITER operator's safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.

EXPERIENCES FROM THE SOURCE-TERM ANALYSIS OF A LOW AND INTERMEDIATE LEVEL RADWASTE DISPOSAL FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park,Jin Beak; Park, Joo-Wan; Lee, Eun-Young

2003-02-27

Enhancement of a computer code SAGE for evaluation of the Korean concept for a LILW waste disposal facility is discussed. Several features of source term analysis are embedded into SAGE to analyze: (1) effects of degradation mode of an engineered barrier, (2) effects of dispersion phenomena in the unsaturated zone and (3) effects of time dependent sorption coefficient in the unsaturated zone. IAEA's Vault Safety Case (VSC) approach is used to demonstrate the ability of this assessment code. Results of MASCOT are used for comparison purposes. These enhancements of the safety assessment code, SAGE, can contribute to realistic evaluation ofmore » the Korean concept of the LILW disposal project in the near future.« less

Modification of the SAS4A Safety Analysis Code for Integration with the ADAPT Discrete Dynamic Event Tree Framework.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary Kyle; Denman, Matthew R.

It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through themore » analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s majormore » emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.« less

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoo, Junsoo; Choi, Yong-joon; Smith, Curtis L.

2016-10-01

The RELAP-7, a safety analysis code for nuclear reactor system, is under development at Idaho National Laboratory (INL). Overall, the code development is directed towards leveraging the advancements in computer science technology, numerical solution methods and physical models over the last decades. Recently, INL has also been putting an effort to establish the code assessment plan, which aims to ensure an improved final product quality through the RELAP-7 development process. The ultimate goal of this plan is to propose a suitable way to systematically assess the wide range of software requirements for RELAP-7, including the software design, user interface, andmore » technical requirements, etc. To this end, we first survey the literature (i.e., international/domestic reports, research articles) addressing the desirable features generally required for advanced nuclear system safety analysis codes. In addition, the V&V (verification and validation) efforts as well as the legacy issues of several recently-developed codes (e.g., RELAP5-3D, TRACE V5.0) are investigated. Lastly, this paper outlines the Requirement Traceability Matrix (RTM) for RELAP-7 which can be used to systematically evaluate and identify the code development process and its present capability.« less

Data Analysis Approaches for the Risk-Informed Safety Margins Characterization Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Alfonsi, Andrea; Maljovec, Daniel P.

2016-09-01

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, RELAP-7, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually calledmore » Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, “extracting information” means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework.« less

Alternative Fuels Data Center: Biodiesel Codes, Standards, and Safety

Science.gov Websites

Codes, Standards, and Safety to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Codes, Standards, and Safety on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Codes , Standards, and Safety on Twitter Bookmark Alternative Fuels Data Center: Biodiesel Codes, Standards, and

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guo, Z.; Zweibaum, N.; Shao, M.

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate amore » staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. In conclusion, this paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Delay Analysis of Car-to-Car Reliable Data Delivery Strategies Based on Data Mulling with Network Coding

NASA Astrophysics Data System (ADS)

Park, Joon-Sang; Lee, Uichin; Oh, Soon Young; Gerla, Mario; Lun, Desmond Siumen; Ro, Won Woo; Park, Joonseok

Vehicular ad hoc networks (VANET) aims to enhance vehicle navigation safety by providing an early warning system: any chance of accidents is informed through the wireless communication between vehicles. For the warning system to work, it is crucial that safety messages be reliably delivered to the target vehicles in a timely manner and thus reliable and timely data dissemination service is the key building block of VANET. Data mulling technique combined with three strategies, network codeing, erasure coding and repetition coding, is proposed for the reliable and timely data dissemination service. Particularly, vehicles in the opposite direction on a highway are exploited as data mules, mobile nodes physically delivering data to destinations, to overcome intermittent network connectivity cause by sparse vehicle traffic. Using analytic models, we show that in such a highway data mulling scenario the network coding based strategy outperforms erasure coding and repetition based strategies.

Verification and Validation in a Rapid Software Development Process

NASA Technical Reports Server (NTRS)

Callahan, John R.; Easterbrook, Steve M.

1997-01-01

The high cost of software production is driving development organizations to adopt more automated design and analysis methods such as rapid prototyping, computer-aided software engineering (CASE) tools, and high-level code generators. Even developers of safety-critical software system have adopted many of these new methods while striving to achieve high levels Of quality and reliability. While these new methods may enhance productivity and quality in many cases, we examine some of the risks involved in the use of new methods in safety-critical contexts. We examine a case study involving the use of a CASE tool that automatically generates code from high-level system designs. We show that while high-level testing on the system structure is highly desirable, significant risks exist in the automatically generated code and in re-validating releases of the generated code after subsequent design changes. We identify these risks and suggest process improvements that retain the advantages of rapid, automated development methods within the quality and reliability contexts of safety-critical projects.

Knowledge management: Role of the the Radiation Safety Information Computational Center (RSICC)

NASA Astrophysics Data System (ADS)

Valentine, Timothy

2017-09-01

The Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL) is an information analysis center that collects, archives, evaluates, synthesizes and distributes information, data and codes that are used in various nuclear technology applications. RSICC retains more than 2,000 software packages that have been provided by code developers from various federal and international agencies. RSICC's customers (scientists, engineers, and students from around the world) obtain access to such computing codes (source and/or executable versions) and processed nuclear data files to promote on-going research, to ensure nuclear and radiological safety, and to advance nuclear technology. The role of such information analysis centers is critical for supporting and sustaining nuclear education and training programs both domestically and internationally, as the majority of RSICC's customers are students attending U.S. universities. Additionally, RSICC operates a secure CLOUD computing system to provide access to sensitive export-controlled modeling and simulation (M&S) tools that support both domestic and international activities. This presentation will provide a general review of RSICC's activities, services, and systems that support knowledge management and education and training in the nuclear field.

Current and anticipated uses of thermal hydraulic codes in Korea

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kyung-Doo; Chang, Won-Pyo

1997-07-01

In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codesmore » with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.« less

38 CFR 61.20 - Life Safety Code capital grants.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 38 Pensions, Bonuses, and Veterans' Relief 2 2011-07-01 2011-07-01 false Life Safety Code capital... (CONTINUED) VA HOMELESS PROVIDERS GRANT AND PER DIEM PROGRAM § 61.20 Life Safety Code capital grants. (a) This section sets forth provisions for obtaining a Life Safety Code capital grant under 38 U.S.C. 2012...

38 CFR 61.20 - Life Safety Code capital grants.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 38 Pensions, Bonuses, and Veterans' Relief 2 2010-07-01 2010-07-01 false Life Safety Code capital... (CONTINUED) VA HOMELESS PROVIDERS GRANT AND PER DIEM PROGRAM § 61.20 Life Safety Code capital grants. (a) This section sets forth provisions for obtaining a Life Safety Code capital grant under 38 U.S.C. 2012...

38 CFR 61.20 - Life Safety Code capital grants.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 38 Pensions, Bonuses, and Veterans' Relief 2 2012-07-01 2012-07-01 false Life Safety Code capital... (CONTINUED) VA HOMELESS PROVIDERS GRANT AND PER DIEM PROGRAM § 61.20 Life Safety Code capital grants. (a) This section sets forth provisions for obtaining a Life Safety Code capital grant under 38 U.S.C. 2012...

Generating Customized Verifiers for Automatically Generated Code

NASA Technical Reports Server (NTRS)

Denney, Ewen; Fischer, Bernd

2008-01-01

Program verification using Hoare-style techniques requires many logical annotations. We have previously developed a generic annotation inference algorithm that weaves in all annotations required to certify safety properties for automatically generated code. It uses patterns to capture generator- and property-specific code idioms and property-specific meta-program fragments to construct the annotations. The algorithm is customized by specifying the code patterns and integrating them with the meta-program fragments for annotation construction. However, this is difficult since it involves tedious and error-prone low-level term manipulations. Here, we describe an annotation schema compiler that largely automates this customization task using generative techniques. It takes a collection of high-level declarative annotation schemas tailored towards a specific code generator and safety property, and generates all customized analysis functions and glue code required for interfacing with the generic algorithm core, thus effectively creating a customized annotation inference algorithm. The compiler raises the level of abstraction and simplifies schema development and maintenance. It also takes care of some more routine aspects of formulating patterns and schemas, in particular handling of irrelevant program fragments and irrelevant variance in the program structure, which reduces the size, complexity, and number of different patterns and annotation schemas that are required. The improvements described here make it easier and faster to customize the system to a new safety property or a new generator, and we demonstrate this by customizing it to certify frame safety of space flight navigation code that was automatically generated from Simulink models by MathWorks' Real-Time Workshop.

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Cao, L.; Ohkawa, K.

2012-07-01

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model ismore » conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)« less

PCC Framework for Program-Generators

NASA Technical Reports Server (NTRS)

Kong, Soonho; Choi, Wontae; Yi, Kwangkeun

2009-01-01

In this paper, we propose a proof-carrying code framework for program-generators. The enabling technique is abstract parsing, a static string analysis technique, which is used as a component for generating and validating certificates. Our framework provides an efficient solution for certifying program-generators whose safety properties are expressed in terms of the grammar representing the generated program. The fixed-point solution of the analysis is generated and attached with the program-generator on the code producer side. The consumer receives the code with a fixed-point solution and validates that the received fixed point is indeed a fixed point of the received code. This validation can be done in a single pass.

Synthesizing Safety Conditions for Code Certification Using Meta-Level Programming

NASA Technical Reports Server (NTRS)

Eusterbrock, Jutta

2004-01-01

In code certification the code consumer publishes a safety policy and the code producer generates a proof that the produced code is in compliance with the published safety policy. In this paper, a novel viewpoint approach towards an implementational re-use oriented framework for code certification is taken. It adopts ingredients from Necula's approach for proof-carrying code, but in this work safety properties can be analyzed on a higher code level than assembly language instructions. It consists of three parts: (1) The specification language is extended to include generic pre-conditions that shall ensure safety at all states that can be reached during program execution. Actual safety requirements can be expressed by providing domain-specific definitions for the generic predicates which act as interface to the environment. (2) The Floyd-Hoare inductive assertion method is refined to obtain proof rules that allow the derivation of the proof obligations in terms of the generic safety predicates. (3) A meta-interpreter is designed and experimentally implemented that enables automatic synthesis of proof obligations for submitted programs by applying the modified Floyd-Hoare rules. The proof obligations have two separate conjuncts, one for functional correctness and another for the generic safety obligations. Proof of the generic obligations, having provided the actual safety definitions as context, ensures domain-specific safety of program execution in a particular environment and is simpler than full program verification.

Analysis of the technology acceptance model in examining hospital nurses' behavioral intentions toward the use of bar code medication administration.

PubMed

Song, Lunar; Park, Byeonghwa; Oh, Kyeung Mi

2015-04-01

Serious medication errors continue to exist in hospitals, even though there is technology that could potentially eliminate them such as bar code medication administration. Little is known about the degree to which the culture of patient safety is associated with behavioral intention to use bar code medication administration. Based on the Technology Acceptance Model, this study evaluated the relationships among patient safety culture and perceived usefulness and perceived ease of use, and behavioral intention to use bar code medication administration technology among nurses in hospitals. Cross-sectional surveys with a convenience sample of 163 nurses using bar code medication administration were conducted. Feedback and communication about errors had a positive impact in predicting perceived usefulness (β=.26, P<.01) and perceived ease of use (β=.22, P<.05). In a multiple regression model predicting for behavioral intention, age had a negative impact (β=-.17, P<.05); however, teamwork within hospital units (β=.20, P<.05) and perceived usefulness (β=.35, P<.01) both had a positive impact on behavioral intention. The overall bar code medication administration behavioral intention model explained 24% (P<.001) of the variance. Identified factors influencing bar code medication administration behavioral intention can help inform hospitals to develop tailored interventions for RNs to reduce medication administration errors and increase patient safety by using this technology.

Residential building codes, affordability, and health protection: a risk-tradeoff approach.

PubMed

Hammitt, J K; Belsky, E S; Levy, J I; Graham, J D

1999-12-01

Residential building codes intended to promote health and safety may produce unintended countervailing risks by adding to the cost of construction. Higher construction costs increase the price of new homes and may increase health and safety risks through "income" and "stock" effects. The income effect arises because households that purchase a new home have less income remaining for spending on other goods that contribute to health and safety. The stock effect arises because suppression of new-home construction leads to slower replacement of less safe housing units. These countervailing risks are not presently considered in code debates. We demonstrate the feasibility of estimating the approximate magnitude of countervailing risks by combining the income effect with three relatively well understood and significant home-health risks. We estimate that a code change that increases the nationwide cost of constructing and maintaining homes by $150 (0.1% of the average cost to build a single-family home) would induce offsetting risks yielding between 2 and 60 premature fatalities or, including morbidity effects, between 20 and 800 lost quality-adjusted life years (both discounted at 3%) each year the code provision remains in effect. To provide a net health benefit, the code change would need to reduce risk by at least this amount. Future research should refine these estimates, incorporate quantitative uncertainty analysis, and apply a full risk-tradeoff approach to real-world case studies of proposed code changes.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

SCALE Code System 6.2.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parisi, Carlo; Prescott, Steve; Ma, Zhegang

This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA modelsmore » for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.« less

Perception of School Safety of a Local School

ERIC Educational Resources Information Center

Massey-Jones, Darla

2013-01-01

This qualitative case study investigated the perception of school safety, what current policies and procedures were effective, and what policies and procedures should be implemented. Data were collected in two steps, by survey and focus group interview. Analysis determined codes that revealed several themes relevant to the perception of school…

Comparing the coding of complications in Queensland and Victorian admitted patient data.

PubMed

Michel, Jude L; Cheng, Diana; Jackson, Terri J

2011-08-01

To examine differences between Queensland and Victorian coding of hospital-acquired conditions and suggest ways to improve the usefulness of these data in the monitoring of patient safety events. Secondary analysis of admitted patient episode data collected in Queensland and Victoria. Comparison of depth of coding, and patterns in the coding of ten commonly coded complications of five elective procedures. Comparison of the mean complication codes assigned per episode revealed Victoria assigns more valid codes than Queensland for all procedures, with the difference between the states being significantly different in all cases. The proportion of the codes flagged as complications was consistently lower for Queensland when comparing 10 common complications for each of the five selected elective procedures. The estimated complication rates for the five procedures showed Victoria to have an apparently higher complication rate than Queensland for 35 of the 50 complications examined. Our findings demonstrate that the coding of complications is more comprehensive in Victoria than in Queensland. It is known that inconsistencies exist between states in routine hospital data quality. Comparative use of patient safety indicators should be viewed with caution until standards are improved across Australia. More exploration of data quality issues is needed to identify areas for improvement.

Response surface method in geotechnical/structural analysis, phase 1

NASA Astrophysics Data System (ADS)

Wong, F. S.

1981-02-01

In the response surface approach, an approximating function is fit to a long running computer code based on a limited number of code calculations. The approximating function, called the response surface, is then used to replace the code in subsequent repetitive computations required in a statistical analysis. The procedure of the response surface development and feasibility of the method are shown using a sample problem in slop stability which is based on data from centrifuge experiments of model soil slopes and involves five random soil parameters. It is shown that a response surface can be constructed based on as few as four code calculations and that the response surface is computationally extremely efficient compared to the code calculation. Potential applications of this research include probabilistic analysis of dynamic, complex, nonlinear soil/structure systems such as slope stability, liquefaction, and nuclear reactor safety.

76 FR 11339 - Update to NFPA 101, Life Safety Code, for State Home Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-02

... DEPARTMENT OF VETERANS AFFAIRS 38 CFR Part 51 RIN 2900-AN59 Update to NFPA 101, Life Safety Code..., Life Safety Code. The change is designed to assure that State Home facilities meet current industry- wide standards regarding life safety and fire safety. DATES: Effective Date: This final rule is...

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

29 CFR 1915.90 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 29 Labor 7 2013-07-01 2013-07-01 false Safety color code for marking physical hazards. 1915.90 Section 1915.90 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH... General Working Conditions § 1915.90 Safety color code for marking physical hazards. The requirements...

29 CFR 1915.90 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 29 Labor 7 2014-07-01 2014-07-01 false Safety color code for marking physical hazards. 1915.90 Section 1915.90 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH... General Working Conditions § 1915.90 Safety color code for marking physical hazards. The requirements...

29 CFR 1915.90 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 29 Labor 7 2012-07-01 2012-07-01 false Safety color code for marking physical hazards. 1915.90 Section 1915.90 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH... General Working Conditions § 1915.90 Safety color code for marking physical hazards. The requirements...

Aviation Safety Modeling and Simulation (ASMM) Propulsion Fleet Modeling: A Tool for Semi-Automatic Construction of CORBA-based Applications from Legacy Fortran Programs

NASA Technical Reports Server (NTRS)

Sang, Janche

2003-01-01

Within NASA's Aviation Safety Program, NASA GRC participates in the Modeling and Simulation Project called ASMM. NASA GRC s focus is to characterize the propulsion systems performance from a fleet management and maintenance perspective by modeling and through simulation predict the characteristics of two classes of commercial engines (CFM56 and GE90). In prior years, the High Performance Computing and Communication (HPCC) program funded, NASA Glenn in developing a large scale, detailed simulations for the analysis and design of aircraft engines called the Numerical Propulsion System Simulation (NPSS). Three major aspects of this modeling included the integration of different engine components, coupling of multiple disciplines, and engine component zooming at appropriate level fidelity, require relatively tight coupling of different analysis codes. Most of these codes in aerodynamics and solid mechanics are written in Fortran. Refitting these legacy Fortran codes with distributed objects can increase these codes reusability. Aviation Safety s modeling and simulation use in characterizing fleet management has similar needs. The modeling and simulation of these propulsion systems use existing Fortran and C codes that are instrumental in determining the performance of the fleet. The research centers on building a CORBA-based development environment for programmers to easily wrap and couple legacy Fortran codes. This environment consists of a C++ wrapper library to hide the details of CORBA and an efficient remote variable scheme to facilitate data exchange between the client and the server model. Additionally, a Web Service model should also be constructed for evaluation of this technology s use over the next two- three years.

76 FR 77549 - Colorado River Indian Tribes-Amendment to Health & Safety Code, Article 2. Liquor

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-13

... Health & Safety Code, Article 2. Liquor AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice. SUMMARY: This notice publishes the amendment to the Colorado River Tribal Health and Safety Code, Article... Code, Article 2, Liquor by Ordinance No. 10-03 on December 13, 2010. This notice is published in...

SAFETY IN THE DESIGN OF SCIENCE LABORATORIES AND BUILDING CODES.

ERIC Educational Resources Information Center

HOROWITZ, HAROLD

THE DESIGN OF COLLEGE AND UNIVERSITY BUILDINGS USED FOR SCIENTIFIC RESEARCH AND EDUCATION IS DISCUSSED IN TERMS OF LABORATORY SAFETY AND BUILDING CODES AND REGULATIONS. MAJOR TOPIC AREAS ARE--(1) SAFETY RELATED DESIGN FEATURES OF SCIENCE LABORATORIES, (2) LABORATORY SAFETY AND BUILDING CODES, AND (3) EVIDENCE OF UNSAFE DESIGN. EXAMPLES EMPHASIZE…

Highway Safety Program Manual: Volume 6: Codes and Laws.

ERIC Educational Resources Information Center

National Highway Traffic Safety Administration (DOT), Washington, DC.

Volume 6 of the 19-volume Highway Safety Program Manual (which provides guidance to State and local governments on preferred safety practices) concentrates on codes and laws. The purpose and specific objectives of the Codes and Laws Program, Federal authority in the area of highway safety, and policies regarding traffic regulation are described.…

Security analysis of boolean algebra based on Zhang-Wang digital signature scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zheng, Jinbin, E-mail: jbzheng518@163.com

2014-10-06

In 2005, Zhang and Wang proposed an improvement signature scheme without using one-way hash function and message redundancy. In this paper, we show that this scheme exits potential safety concerns through the analysis of boolean algebra, such as bitwise exclusive-or, and point out that mapping is not one to one between assembly instructions and machine code actually by means of the analysis of the result of the assembly program segment, and which possibly causes safety problems unknown to the software.

Probabilistic margin evaluation on accidental transients for the ASTRID reactor project

NASA Astrophysics Data System (ADS)

Marquès, Michel

2014-06-01

ASTRID is a technological demonstrator of Sodium cooled Fast Reactor (SFR) under development. The conceptual design studies are being conducted in accordance with the Generation IV reactor objectives, particularly in terms of improving safety. For the hypothetical events, belonging to the accidental category "severe accident prevention situations" having a very low frequency of occurrence, the safety demonstration is no more based on a deterministic demonstration with conservative assumptions on models and parameters but on a "Best-Estimate Plus Uncertainty" (BEPU) approach. This BEPU approach ispresented in this paper for an Unprotected Loss-of-Flow (ULOF) event. The Best-Estimate (BE) analysis of this ULOFt ransient is performed with the CATHARE2 code, which is the French reference system code for SFR applications. The objective of the BEPU analysis is twofold: first evaluate the safety margin to sodium boiling in taking into account the uncertainties on the input parameters of the CATHARE2 code (twenty-two uncertain input parameters have been identified, which can be classified into five groups: reactor power, accident management, pumps characteristics, reactivity coefficients, thermal parameters and head losses); secondly quantify the contribution of each input uncertainty to the overall uncertainty of the safety margins, in order to refocusing R&D efforts on the most influential factors. This paper focuses on the methodological aspects of the evaluation of the safety margin. At least for the preliminary phase of the project (conceptual design), a probabilistic criterion has been fixed in the context of this BEPU analysis; this criterion is the value of the margin to sodium boiling, which has a probability 95% to be exceeded, obtained with a confidence level of 95% (i.e. the M5,95percentile of the margin distribution). This paper presents two methods used to assess this percentile: the Wilks method and the Bootstrap method ; the effectiveness of the two methods is compared on the basis of 500 simulations performed with theCATHARE2 code. We conclude that, with only 100 simulations performed with the CATHARE2 code, which is a number of simulations workable in the conceptual design phase of the ASTRID project where the models and the hypothesis are often modified, it is best in order to evaluate the percentile M5,95 of the margin to sodium boiling to use the bootstrap method, which will provide a slightly conservative result. On the other hand, in order to obtain an accurate estimation of the percentileM5,95, for the safety report for example, it will be necessary to perform at least 300 simulations with the CATHARE2 code. In this case, both methods (Wilks and Bootstrap) would give equivalent results.

Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arndt, S.A.

1997-07-01

The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for codemore » use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

RAVEN Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alfonsi, Andrea; Rabiti, Cristian; Mandelli, Diego

2016-06-01

RAVEN is a software framework able to perform parametric and stochastic analysis based on the response of complex system codes. The initial development was aimed at providing dynamic risk analysis capabilities to the thermohydraulic code RELAP-7, currently under development at Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose stochastic and uncertainty quantification platform, capable of communicating with any system code. In fact, the provided Application Programming Interfaces (APIs) allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by input filesmore » or via python interfaces. RAVEN is capable of investigating system response and explore input space using various sampling schemes such as Monte Carlo, grid, or Latin hypercube. However, RAVEN strength lies in its system feature discovery capabilities such as: constructing limit surfaces, separating regions of the input space leading to system failure, and using dynamic supervised learning techniques. The development of RAVEN started in 2012 when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework arose. RAVEN’s principal assignment is to provide the necessary software and algorithms in order to employ the concepts developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just to identify the frequency of an event potentially leading to a system failure, but the proximity (or lack thereof) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. peak pressure in a pipe) is exceeded under certain conditions. Most of the capabilities, implemented having RELAP-7 as a principal focus, are easily deployable to other system codes. For this reason, several side activates have been employed (e.g. RELAP5-3D, any MOOSE-based App, etc.) or are currently ongoing for coupling RAVEN with several different software. The aim of this document is to provide a set of commented examples that can help the user to become familiar with the RAVEN code usage.« less

41 CFR 128-1.8005 - Seismic safety standards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... the model building codes that the Interagency Committee on Seismic Safety in Construction (ICSSC...) Uniform Building Code (UBC); (2) The 1992 Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code (NBC); and (3) The 1992 Amendments to the Southern Building Code...

38 CFR 39.63 - Architectural design standards.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 5000, Building Construction and Safety Code, and the 2002 edition of the National Electrical Code, NFPA... 5000, Building Construction and Safety Code. Where the adopted codes state conflicting requirements... the standards set forth in this section, all applicable local and State building codes and regulations...

41 CFR 128-1.8005 - Seismic safety standards.

Code of Federal Regulations, 2012 CFR

2012-01-01

... the model building codes that the Interagency Committee on Seismic Safety in Construction (ICSSC...) Uniform Building Code (UBC); (2) The 1992 Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code (NBC); and (3) The 1992 Amendments to the Southern Building Code...

38 CFR 39.63 - Architectural design standards.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 5000, Building Construction and Safety Code, and the 2002 edition of the National Electrical Code, NFPA... 5000, Building Construction and Safety Code. Where the adopted codes state conflicting requirements... the standards set forth in this section, all applicable local and State building codes and regulations...

38 CFR 39.63 - Architectural design standards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 5000, Building Construction and Safety Code, and the 2002 edition of the National Electrical Code, NFPA... 5000, Building Construction and Safety Code. Where the adopted codes state conflicting requirements... the standards set forth in this section, all applicable local and State building codes and regulations...

41 CFR 128-1.8005 - Seismic safety standards.

Code of Federal Regulations, 2014 CFR

2014-01-01

... the model building codes that the Interagency Committee on Seismic Safety in Construction (ICSSC...) Uniform Building Code (UBC); (2) The 1992 Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code (NBC); and (3) The 1992 Amendments to the Southern Building Code...

41 CFR 128-1.8005 - Seismic safety standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the model building codes that the Interagency Committee on Seismic Safety in Construction (ICSSC...) Uniform Building Code (UBC); (2) The 1992 Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code (NBC); and (3) The 1992 Amendments to the Southern Building Code...

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

Creation of the Naturalistic Engagement in Secondary Tasks (NEST) distracted driving dataset.

PubMed

Owens, Justin M; Angell, Linda; Hankey, Jonathan M; Foley, James; Ebe, Kazutoshi

2015-09-01

Distracted driving has become a topic of critical importance to driving safety research over the past several decades. Naturalistic driving data offer a unique opportunity to study how drivers engage with secondary tasks in real-world driving; however, the complexities involved with identifying and coding relevant epochs of naturalistic data have limited its accessibility to the general research community. This project was developed to help address this problem by creating an accessible dataset of driver behavior and situational factors observed during distraction-related safety-critical events and baseline driving epochs, using the Strategic Highway Research Program 2 (SHRP2) naturalistic dataset. The new NEST (Naturalistic Engagement in Secondary Tasks) dataset was created using crashes and near-crashes from the SHRP2 dataset that were identified as including secondary task engagement as a potential contributing factor. Data coding included frame-by-frame video analysis of secondary task and hands-on-wheel activity, as well as summary event information. In addition, information about each secondary task engagement within the trip prior to the crash/near-crash was coded at a higher level. Data were also coded for four baseline epochs and trips per safety-critical event. 1,180 events and baseline epochs were coded, and a dataset was constructed. The project team is currently working to determine the most useful way to allow broad public access to the dataset. We anticipate that the NEST dataset will be extraordinarily useful in allowing qualified researchers access to timely, real-world data concerning how drivers interact with secondary tasks during safety-critical events and baseline driving. The coded dataset developed for this project will allow future researchers to have access to detailed data on driver secondary task engagement in the real world. It will be useful for standalone research, as well as for integration with additional SHRP2 data to enable the conduct of more complex research. Copyright © 2015 Elsevier Ltd and National Safety Council. All rights reserved.

Heat Transfer Principles in Thermal Calculation of Structures in Fire

PubMed Central

Zhang, Chao; Usmani, Asif

2016-01-01

Structural fire engineering (SFE) is a relatively new interdisciplinary subject, which requires a comprehensive knowledge of heat transfer, fire dynamics and structural analysis. It is predominantly the community of structural engineers who currently carry out most of the structural fire engineering research and design work. The structural engineering curriculum in universities and colleges do not usually include courses in heat transfer and fire dynamics. In some institutions of higher education, there are graduate courses for fire resistant design which focus on the design approaches in codes. As a result, structural engineers who are responsible for structural fire safety and are competent to do their jobs by following the rules specified in prescriptive codes may find it difficult to move toward performance-based fire safety design which requires a deep understanding of both fire and heat. Fire safety engineers, on the other hand, are usually focused on fire development and smoke control, and may not be familiar with the heat transfer principles used in structural fire analysis, or structural failure analysis. This paper discusses the fundamental heat transfer principles in thermal calculation of structures in fire, which might serve as an educational guide for students, engineers and researchers. Insights on problems which are commonly ignored in performance based fire safety design are also presented. PMID:26783379

Technology Infusion of CodeSonar into the Space Network Ground Segment (RII07)

NASA Technical Reports Server (NTRS)

Benson, Markland

2008-01-01

The NASA Software Assurance Research Program (in part) performs studies as to the feasibility of technologies for improving the safety, quality, reliability, cost, and performance of NASA software. This study considers the application of commercial automated source code analysis tools to mission critical ground software that is in the operations and sustainment portion of the product lifecycle.

Alternative Fuels Data Center: E85 Codes and Standards

Science.gov Websites

Development Equipment Options Equipment Installation Codes, Standards, & Safety Vehicles Laws & ; Incentives Ethanol Codes, Standards, and Safety The U.S. Environmental Protection Agency's (EPA) Office of -Gasoline Blends. The Occupational Safety and Health Administration (OSHA) regulates some fuel-dispensing

7 CFR 1724.50 - Compliance with National Electrical Safety Code (NESC).

Code of Federal Regulations, 2013 CFR

2013-01-01

... 7 Agriculture 11 2013-01-01 2013-01-01 false Compliance with National Electrical Safety Code (NESC... UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE ELECTRIC ENGINEERING, ARCHITECTURAL SERVICES AND DESIGN POLICIES AND PROCEDURES Electric System Design § 1724.50 Compliance with National Electrical Safety Code...

7 CFR 1724.50 - Compliance with National Electrical Safety Code (NESC).

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 11 2010-01-01 2010-01-01 false Compliance with National Electrical Safety Code (NESC... UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE ELECTRIC ENGINEERING, ARCHITECTURAL SERVICES AND DESIGN POLICIES AND PROCEDURES Electric System Design § 1724.50 Compliance with National Electrical Safety Code...

7 CFR 1724.50 - Compliance with National Electrical Safety Code (NESC).

Code of Federal Regulations, 2011 CFR

2011-01-01

... 7 Agriculture 11 2011-01-01 2011-01-01 false Compliance with National Electrical Safety Code (NESC... UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE ELECTRIC ENGINEERING, ARCHITECTURAL SERVICES AND DESIGN POLICIES AND PROCEDURES Electric System Design § 1724.50 Compliance with National Electrical Safety Code...

7 CFR 1724.50 - Compliance with National Electrical Safety Code (NESC).

Code of Federal Regulations, 2012 CFR

2012-01-01

... 7 Agriculture 11 2012-01-01 2012-01-01 false Compliance with National Electrical Safety Code (NESC... UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE ELECTRIC ENGINEERING, ARCHITECTURAL SERVICES AND DESIGN POLICIES AND PROCEDURES Electric System Design § 1724.50 Compliance with National Electrical Safety Code...

7 CFR 1724.50 - Compliance with National Electrical Safety Code (NESC).

Code of Federal Regulations, 2014 CFR

2014-01-01

... 7 Agriculture 11 2014-01-01 2014-01-01 false Compliance with National Electrical Safety Code (NESC... UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE ELECTRIC ENGINEERING, ARCHITECTURAL SERVICES AND DESIGN POLICIES AND PROCEDURES Electric System Design § 1724.50 Compliance with National Electrical Safety Code...

Police accident report forms: safety device coding and enacted laws.

PubMed

Brock, K; Lapidus, G

2008-12-01

Safety device coding on state police accident report (PAR) forms was compared with provisions in state traffic safety laws. PAR forms were obtained from all 50 states and the District of Columbia (states/DC). For seat belts, 22 states/DC had a primary seat belt enforcement law vs 50 with a PAR code. For car seats, all 51 states/DC had a law and a PAR code. For booster seats, 39 states/DC had a law vs nine with a PAR code. For motorcycle helmets, 21 states/DC had an all-age rider helmet law and another 26 a partial-age law vs 50 with a PAR code. For bicycle helmets, 21 states/DC had a partial-age rider helmet law vs 48 with a PAR code. Therefore gaps in the ability of states to fully record accident data reflective of existing state traffic safety laws are revealed. Revising the PAR forms in all states to include complete variables for safety devices should be an important priority, independent of the laws.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs) . Volume 2; Appendices

NASA Technical Reports Server (NTRS)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This document contains the appendices to the main report.

A Comparative Study on Seismic Analysis of Bangladesh National Building Code (BNBC) with Other Building Codes

NASA Astrophysics Data System (ADS)

Bari, Md. S.; Das, T.

2013-09-01

Tectonic framework of Bangladesh and adjoining areas indicate that Bangladesh lies well within an active seismic zone. The after effect of earthquake is more severe in an underdeveloped and a densely populated country like ours than any other developed countries. Bangladesh National Building Code (BNBC) was first established in 1993 to provide guidelines for design and construction of new structure subject to earthquake ground motions in order to minimize the risk to life for all structures. A revision of BNBC 1993 is undergoing to make this up to date with other international building codes. This paper aims at the comparison of various provisions of seismic analysis as given in building codes of different countries. This comparison will give an idea regarding where our country stands when it comes to safety against earth quake. Primarily, various seismic parameters in BNBC 2010 (draft) have been studied and compared with that of BNBC 1993. Later, both 1993 and 2010 edition of BNBC codes have been compared graphically with building codes of other countries such as National Building Code of India 2005 (NBC-India 2005), American Society of Civil Engineering 7-05 (ASCE 7-05). The base shear/weight ratios have been plotted against the height of the building. The investigation in this paper reveals that BNBC 1993 has the least base shear among all the codes. Factored Base shear values of BNBC 2010 are found to have increased significantly than that of BNBC 1993 for low rise buildings (≤20 m) around the country than its predecessor. Despite revision of the code, BNBC 2010 (draft) still suggests less base shear values when compared to the Indian and American code. Therefore, this increase in factor of safety against the earthquake imposed by the proposed BNBC 2010 code by suggesting higher values of base shear is appreciable.

Recent improvements of reactor physics codes in MHI

NASA Astrophysics Data System (ADS)

Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

2015-12-01

This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

Recent improvements of reactor physics codes in MHI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki

2015-12-31

This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipatedmore » transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.« less

29 CFR 1910.144 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 29 Labor 5 2013-07-01 2013-07-01 false Safety color code for marking physical hazards. 1910.144... § 1910.144 Safety color code for marking physical hazards. (a) Color identification—(1) Red. Red shall be the basic color for the identification of: (i) Fire protection equipment and apparatus. [Reserved] (ii...

29 CFR 1910.144 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 29 Labor 5 2014-07-01 2014-07-01 false Safety color code for marking physical hazards. 1910.144... § 1910.144 Safety color code for marking physical hazards. (a) Color identification—(1) Red. Red shall be the basic color for the identification of: (i) Fire protection equipment and apparatus. [Reserved] (ii...

29 CFR 1910.144 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 29 Labor 5 2012-07-01 2012-07-01 false Safety color code for marking physical hazards. 1910.144... § 1910.144 Safety color code for marking physical hazards. (a) Color identification—(1) Red. Red shall be the basic color for the identification of: (i) Fire protection equipment and apparatus. [Reserved] (ii...

29 CFR 1910.144 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 29 Labor 5 2010-07-01 2010-07-01 false Safety color code for marking physical hazards. 1910.144... § 1910.144 Safety color code for marking physical hazards. (a) Color identification—(1) Red. Red shall be... basic color for designating caution and for marking physical hazards such as: Striking against...

30 CFR 905.773 - Requirements for permits and permit processing.

Code of Federal Regulations, 2010 CFR

2010-07-01

..., 42 U.S.C. 7401 et seq California Air Pollution Control Laws, Cal. Health & Safety Code section 39000... (11) Noise Control Act, 42 U.S.C. 4903 California Noise Control Act of 1973, Cal. Health & Safety Code... Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the Hazardous Waste Control Law...

26 CFR 1.42-5 - Monitoring compliance with low-income housing credit requirements.

Code of Federal Regulations, 2012 CFR

2012-04-01

... be required to retain the original local health, safety, or building code violation reports or... account local health, safety, and building codes (or other habitability standards), and the State or local government unit responsible for making local health, safety, or building code inspections did not issue a...

26 CFR 1.42-5 - Monitoring compliance with low-income housing credit requirements.

Code of Federal Regulations, 2010 CFR

2010-04-01

... be required to retain the original local health, safety, or building code violation reports or... account local health, safety, and building codes (or other habitability standards), and the State or local government unit responsible for making local health, safety, or building code inspections did not issue a...

Fault trees for decision making in systems analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lambert, Howard E.

1975-10-09

The application of fault tree analysis (FTA) to system safety and reliability is presented within the framework of system safety analysis. The concepts and techniques involved in manual and automated fault tree construction are described and their differences noted. The theory of mathematical reliability pertinent to FTA is presented with emphasis on engineering applications. An outline of the quantitative reliability techniques of the Reactor Safety Study is given. Concepts of probabilistic importance are presented within the fault tree framework and applied to the areas of system design, diagnosis and simulation. The computer code IMPORTANCE ranks basic events and cut setsmore » according to a sensitivity analysis. A useful feature of the IMPORTANCE code is that it can accept relative failure data as input. The output of the IMPORTANCE code can assist an analyst in finding weaknesses in system design and operation, suggest the most optimal course of system upgrade, and determine the optimal location of sensors within a system. A general simulation model of system failure in terms of fault tree logic is described. The model is intended for efficient diagnosis of the causes of system failure in the event of a system breakdown. It can also be used to assist an operator in making decisions under a time constraint regarding the future course of operations. The model is well suited for computer implementation. New results incorporated in the simulation model include an algorithm to generate repair checklists on the basis of fault tree logic and a one-step-ahead optimization procedure that minimizes the expected time to diagnose system failure.« less

Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Page, R.; Jones, J.R.

1997-07-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation toolsmore » is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.« less

Use of the Means/Ends Test to Evaluate Public School Dress-Code Policies

ERIC Educational Resources Information Center

Workman, Jane E.; Studak, Cathryn M.

2008-01-01

The purpose of this article is to explain how a means/ends test can be adapted for the school environment. Public school officials can use a means/ends test to document an analysis of whether dress-code policies will be effective in diminishing risks to the health, safety, or morality of the school population. Elements of policy evaluation--ends,…

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

NASA Astrophysics Data System (ADS)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Runtime Detection of C-Style Errors in UPC Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pirkelbauer, P; Liao, C; Panas, T

2011-09-29

Unified Parallel C (UPC) extends the C programming language (ISO C 99) with explicit parallel programming support for the partitioned global address space (PGAS), which provides a global memory space with localized partitions to each thread. Like its ancestor C, UPC is a low-level language that emphasizes code efficiency over safety. The absence of dynamic (and static) safety checks allows programmer oversights and software flaws that can be hard to spot. In this paper, we present an extension of a dynamic analysis tool, ROSE-Code Instrumentation and Runtime Monitor (ROSECIRM), for UPC to help programmers find C-style errors involving the globalmore » address space. Built on top of the ROSE source-to-source compiler infrastructure, the tool instruments source files with code that monitors operations and keeps track of changes to the system state. The resulting code is linked to a runtime monitor that observes the program execution and finds software defects. We describe the extensions to ROSE-CIRM that were necessary to support UPC. We discuss complications that arise from parallel code and our solutions. We test ROSE-CIRM against a runtime error detection test suite, and present performance results obtained from running error-free codes. ROSE-CIRM is released as part of the ROSE compiler under a BSD-style open source license.« less

Java Source Code Analysis for API Migration to Embedded Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Winter, Victor; McCoy, James A.; Guerrero, Jonathan

Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered bymore » APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.« less

SCALE Code System 6.2.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlomore » radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.« less

The construction FACE database - Codifying the NIOSH FACE reports.

PubMed

Dong, Xiuwen Sue; Largay, Julie A; Wang, Xuanwen; Cain, Chris Trahan; Romano, Nancy

2017-09-01

The National Institute for Occupational Safety and Health (NIOSH) has published reports detailing the results of investigations on selected work-related fatalities through the Fatality Assessment and Control Evaluation (FACE) program since 1982. Information from construction-related FACE reports was coded into the Construction FACE Database (CFD). Use of the CFD was illustrated by analyzing major CFD variables. A total of 768 construction fatalities were included in the CFD. Information on decedents, safety training, use of PPE, and FACE recommendations were coded. Analysis shows that one in five decedents in the CFD died within the first two months on the job; 75% and 43% of reports recommended having safety training or installing protection equipment, respectively. Comprehensive research using FACE reports may improve understanding of work-related fatalities and provide much-needed information on injury prevention. The CFD allows researchers to analyze the FACE reports quantitatively and efficiently. Copyright © 2017 Elsevier Ltd and National Safety Council. All rights reserved.

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

29 CFR 1910.119 - Process safety management of highly hazardous chemicals.

Code of Federal Regulations, 2011 CFR

2011-07-01

... complexity of the process will influence the decision as to the appropriate PHA methodology to use. All PHA... process hazard analysis in sufficient detail to support the analysis. (3) Information pertaining to the...) Relief system design and design basis; (E) Ventilation system design; (F) Design codes and standards...

29 CFR 1910.119 - Process safety management of highly hazardous chemicals.

Code of Federal Regulations, 2010 CFR

2010-07-01

... complexity of the process will influence the decision as to the appropriate PHA methodology to use. All PHA... process hazard analysis in sufficient detail to support the analysis. (3) Information pertaining to the...) Relief system design and design basis; (E) Ventilation system design; (F) Design codes and standards...

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs)

NASA Technical Reports Server (NTRS)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs). Corrected Copy, Aug. 25, 2014

NASA Technical Reports Server (NTRS)

Prosser, William H.

2014-01-01

In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

75 FR 17644 - Update to NFPA 101, Life Safety Code, for State Home Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-07

... DEPARTMENT OF VETERANS AFFAIRS 38 CFR Part 51 RIN 2900-AN59 Update to NFPA 101, Life Safety Code... certain provisions of the 2009 edition of the National Fire Protection Association's NFPA 101, Life Safety... standards regarding life safety and fire safety. DATES: Written comments must be received by VA on or before...

RAVEN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph

2015-10-01

RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need tomore » be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an important process such as peak pressure in a pipe) is exceeded under certain conditions. The initial development of RAVEN has been focused on providing dynamic risk assessment capability to RELAP-7, currently under development at the INL and, likely, future replacement of the RELAP5-3D code. Most the capabilities that have been implemented having RELAP-7 as principal focus are easily deployable for other system codes. For this reason, several side activaties are currently ongoing for coupling RAVEN with software such as RELAP5-3D, etc. The aim of this document is the explanation of the input requirements, focalizing on the input structure.« less

RAVEN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph

2016-02-01

RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need tomore » be perturbed are accessible by input files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an important process such as peak pressure in a pipe) is exceeded under certain conditions. The initial development of RAVEN has been focused on providing dynamic risk assessment capability to RELAP-7, currently under development at the INL and, likely, future replacement of the RELAP5-3D code. Most the capabilities that have been implemented having RELAP-7 as principal focus are easily deployable for other system codes. For this reason, several side activates are currently ongoing for coupling RAVEN with software such as RELAP5-3D, etc. The aim of this document is the explanation of the input requirements, focusing on the input structure.« less

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

Kayenta Township Building & Safety Department, Tribal Green Building Code Summit Presentation

EPA Pesticide Factsheets

Tribal Green Building Code Summit Presentation by Kayenta Township Building & Safety Department showing how they established the building department, developed a code adoption and enforcement process, and hired staff to carry out the work.

An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hongbin Zhang; Haihua Zhao; Vincent Mousseau

2008-04-01

Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analysesmore » perspective, we have initiated an effort to develop a high fidelity reactor system safety code.« less

Safety Issues with Hydrogen as a Vehicle Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cadwallader, Lee Charles; Herring, James Stephen

1999-10-01

This report is an initial effort to identify and evaluate safety issues associated with the use of hydrogen as a vehicle fuel in automobiles. Several forms of hydrogen have been considered: gas, liquid, slush, and hydrides. The safety issues have been discussed, beginning with properties of hydrogen and the phenomenology of hydrogen combustion. Safety-related operating experiences with hydrogen vehicles have been summarized to identify concerns that must be addressed in future design activities and to support probabilistic risk assessment. Also, applicable codes, standards, and regulations pertaining to hydrogen usage and refueling have been identified and are briefly discussed. This reportmore » serves as a safety foundation for any future hydrogen safety work, such as a safety analysis or a probabilistic risk assessment.« less

Safety Issues with Hydrogen as a Vehicle Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. C. Cadwallader; J. S. Herring

1999-09-01

This report is an initial effort to identify and evaluate safety issues associated with the use of hydrogen as a vehicle fuel in automobiles. Several forms of hydrogen have been considered: gas, liquid, slush, and hydrides. The safety issues have been discussed, beginning with properties of hydrogen and the phenomenology of hydrogen combustion. Safety-related operating experiences with hydrogen vehicles have been summarized to identify concerns that must be addressed in future design activities and to support probabilistic risk assessment. Also, applicable codes, standards, and regulations pertaining to hydrogen usage and refueling have been identified and are briefly discussed. This reportmore » serves as a safety foundation for any future hydrogen safety work, such as a safety analysis or a probabilistic risk assessment.« less

Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Kula, K.R.; Paik, I.K.; Chung, D.Y.

1996-12-31

Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, includingmore » source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper.« less

Computer codes for checking, plotting and processing of neutron cross-section covariance data and their application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sartori, E.; Roussin, R.W.

This paper presents a brief review of computer codes concerned with checking, plotting, processing and using of covariances of neutron cross-section data. It concentrates on those available from the computer code information centers of the United States and the OECD/Nuclear Energy Agency. Emphasis will be placed also on codes using covariances for specific applications such as uncertainty analysis, data adjustment and data consistency analysis. Recent evaluations contain neutron cross section covariance information for all isotopes of major importance for technological applications of nuclear energy. It is therefore important that the available software tools needed for taking advantage of this informationmore » are widely known as hey permit the determination of better safety margins and allow the optimization of more economic, I designs of nuclear energy systems.« less

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

2014-02-01

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

Seismic Safety Of Simple Masonry Buildings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guadagnuolo, Mariateresa; Faella, Giuseppe

2008-07-08

Several masonry buildings comply with the rules for simple buildings provided by seismic codes. For these buildings explicit safety verifications are not compulsory if specific code rules are fulfilled. In fact it is assumed that their fulfilment ensures a suitable seismic behaviour of buildings and thus adequate safety under earthquakes. Italian and European seismic codes differ in the requirements for simple masonry buildings, mostly concerning the building typology, the building geometry and the acceleration at site. Obviously, a wide percentage of buildings assumed simple by codes should satisfy the numerical safety verification, so that no confusion and uncertainty have tomore » be given rise to designers who must use the codes. This paper aims at evaluating the seismic response of some simple unreinforced masonry buildings that comply with the provisions of the new Italian seismic code. Two-story buildings, having different geometry, are analysed and results from nonlinear static analyses performed by varying the acceleration at site are presented and discussed. Indications on the congruence between code rules and results of numerical analyses performed according to the code itself are supplied and, in this context, the obtained result can provide a contribution for improving the seismic code requirements.« less

RELAP-7 Software Verification and Validation Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less

Fire safety

Treesearch

Robert H. White; Mark A. Dietenberger

1999-01-01

Fire safety is an important concern in all types of construction. The high level of national concern for fire safety is reflected in limitations and design requirements in building codes. These code requirements are discussed in the context of fire safety design and evaluation in the initial section of this chapter. Since basic data on fire behavior of wood products...

29 CFR 1910.35 - Compliance with NFPA 101-2000, Life Safety Code.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 29 Labor 5 2011-07-01 2011-07-01 false Compliance with NFPA 101-2000, Life Safety Code. 1910.35 Section 1910.35 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR OCCUPATIONAL SAFETY AND HEALTH STANDARDS Means of Egress § 1910.35 Compliance...

29 CFR Appendix A to Subpart S of... - References for Further Information

Code of Federal Regulations, 2014 CFR

2014-07-01

... Safety, Health, and Environmental Training. ANSI/IEEE C2-2002 National Electrical Safety Code. ANSI K61.1.... NFPA 59-2004 Utility LP-Gas Plant Code. NFPA 70-2002 National Electrical Code. (See also NFPA 70-2005.... NMAB 353-3-1980 Classification of Combustible Dust in Accordance with the National Electrical Code. [72...

29 CFR Appendix A to Subpart S of... - References for Further Information

Code of Federal Regulations, 2011 CFR

2011-07-01

... Safety, Health, and Environmental Training. ANSI/IEEE C2-2002 National Electrical Safety Code. ANSI K61.1.... NFPA 59-2004 Utility LP-Gas Plant Code. NFPA 70-2002 National Electrical Code. (See also NFPA 70-2005.... NMAB 353-3-1980 Classification of Combustible Dust in Accordance with the National Electrical Code. [72...

29 CFR Appendix A to Subpart S of... - References for Further Information

Code of Federal Regulations, 2012 CFR

2012-07-01

... Safety, Health, and Environmental Training. ANSI/IEEE C2-2002 National Electrical Safety Code. ANSI K61.1.... NFPA 59-2004 Utility LP-Gas Plant Code. NFPA 70-2002 National Electrical Code. (See also NFPA 70-2005.... NMAB 353-3-1980 Classification of Combustible Dust in Accordance with the National Electrical Code. [72...

29 CFR Appendix A to Subpart S of... - References for Further Information

Code of Federal Regulations, 2013 CFR

2013-07-01

... Safety, Health, and Environmental Training. ANSI/IEEE C2-2002 National Electrical Safety Code. ANSI K61.1.... NFPA 59-2004 Utility LP-Gas Plant Code. NFPA 70-2002 National Electrical Code. (See also NFPA 70-2005.... NMAB 353-3-1980 Classification of Combustible Dust in Accordance with the National Electrical Code. [72...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Satyapal, Sunita

The 2011 Annual Progress Report summarizes fiscal year 2011 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; education; market transformation; and systems analysis.

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

The 2013 Annual Progress Report summarizes fiscal year 2013 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

77 FR 42654 - Trifloxystrobin; Pesticide Tolerance

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-20

... code 112). Food manufacturing (NAICS code 311). Pesticide manufacturing (NAICS code 32532). This... filing. III. Aggregate Risk Assessment and Determination of Safety Section 408(b)(2)(A)(i) of FFDCA... dose at which adverse effects of concern are identified (the LOAEL). Uncertainty/safety factors are...

ESAS Deliverable PS 1.1.2.3: Customer Survey on Code Generations in Safety-Critical Applications

NASA Technical Reports Server (NTRS)

Schumann, Johann; Denney, Ewen

2006-01-01

Automated code generators (ACG) are tools that convert a (higher-level) model of a software (sub-)system into executable code without the necessity for a developer to actually implement the code. Although both commercially supported and in-house tools have been used in many industrial applications, little data exists on how these tools are used in safety-critical domains (e.g., spacecraft, aircraft, automotive, nuclear). The aims of the survey, therefore, were threefold: 1) to determine if code generation is primarily used as a tool for prototyping, including design exploration and simulation, or for fiight/production code; 2) to determine the verification issues with code generators relating, in particular, to qualification and certification in safety-critical domains; and 3) to determine perceived gaps in functionality of existing tools.

RELAP-7 Closure Correlations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zou, Ling; Berry, R. A.; Martineau, R. C.

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 codemore » utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.« less

Deriving ICD-10 Codes for Patient Safety Indicators for Large-scale Surveillance Using Administrative Hospital Data.

PubMed

Southern, Danielle A; Burnand, Bernard; Droesler, Saskia E; Flemons, Ward; Forster, Alan J; Gurevich, Yana; Harrison, James; Quan, Hude; Pincus, Harold A; Romano, Patrick S; Sundararajan, Vijaya; Kostanjsek, Nenad; Ghali, William A

2017-03-01

Existing administrative data patient safety indicators (PSIs) have been limited by uncertainty around the timing of onset of included diagnoses. We undertook de novo PSI development through a data-driven approach that drew upon "diagnosis timing" information available in some countries' administrative hospital data. Administrative database analysis and modified Delphi rating process. All hospitalized adults in Canada in 2009. We queried all hospitalizations for ICD-10-CA diagnosis codes arising during hospital stay. We then undertook a modified Delphi panel process to rate the extent to which each of the identified diagnoses has a potential link to suboptimal quality of care. We grouped the identified quality/safety-related diagnoses into relevant clinical categories. Lastly, we queried Alberta hospital discharge data to assess the frequency of the newly defined PSI events. Among 2,416,413 national hospitalizations, we found 2590 unique ICD-10-CA codes flagged as having arisen after admission. Seven panelists evaluated these in a 2-round review process, and identified a listing of 640 ICD-10-CA diagnosis codes judged to be linked to suboptimal quality of care and thus appropriate for inclusion in PSIs. These were then grouped by patient safety experts into 18 clinically relevant PSI categories. We then analyzed data on 2,381,652 Alberta hospital discharges from 2005 through 2012, and found that 134,299 (5.2%) hospitalizations had at least 1 PSI diagnosis. The resulting work creates a foundation for a new set of PSIs for routine large-scale surveillance of hospital and health system performance.

Knowledge of and attitude towards road traffic codes among commercial motorcycle riders in Anambra State.

PubMed

Adogu, O U; Ilika, A L

2006-12-01

Road traffic accidents (rtas) represent a major epidemic of non communicable disease in the country and has since escalated with the introduction of the new phenomenon of commercial motorcycle transportation such as is found in the two urban towns of nnewi and Awka of Anambra state, Nigeria. making use of a pre-tested, semi structured, interviewer administered questionnaire, relevant data on socio demographic and motorcycle characteristics were collected from a sample of commercial motorcyclists selected by systematic sampling technique. their knowledge of and attitude towards road traffic and safety codes were elicited. The result showed that the all-male commercial motorcyclists had a mean age of 30+8.9 years. one hundred and seventy six (32.6%) possessed good knowledge of road traffic codes and safety, while 35 (6.5%) exhibited good attitude towards them. both knowledge of and attitude towards traffic codes and safety improved with increase in educational level (p<0.005, p<0.001 respectively). the younger motorcyclists also possessed statistically significant better knowledge of traffic codes than their older counterparts (p<0.025). attitude to traffic codes and safety had no association with age of the motorcyclists (p>0.25). the study has provided useful information on the knowledge of and attitude towards road traffic and safety codes among commercial motorcyclists in nigeria. pursuit of knowledge through formal and informal education should run pari pasu with efforts to improve the nigerian economy in order to ensure a sustainable positive attitudinal change towards road traffic codes and safety among commercial motorcyclists.

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Kombucha brewing under the Food and Drug Administration model Food Code: risk analysis and processing guidance.

PubMed

Nummer, Brian A

2013-11-01

Kombucha is a fermented beverage made from brewed tea and sugar. The taste is slightly sweet and acidic and it may have residual carbon dioxide. Kombucha is consumed in many countries as a health beverage and it is gaining in popularity in the U.S. Consequently, many retailers and food service operators are seeking to brew this beverage on site. As a fermented beverage, kombucha would be categorized in the Food and Drug Administration model Food Code as a specialized process and would require a variance with submission of a food safety plan. This special report was created to assist both operators and regulators in preparing or reviewing a kombucha food safety plan.

Fire safety of wood construction

Treesearch

Robert H. White; Mark A. Dietenberger

2010-01-01

Fire safety is an important concern in all types of construction. The high level of national concern for fire safety is reflected in limitations and design requirements in building codes. These code requirements and related fire performance data are discussed in the context of fire safety design and evaluation in the initial section of this chapter. Because basic data...

Items Supporting the Hanford Internal Dosimetry Program Implementation of the IMBA Computer Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbaugh, Eugene H.; Bihl, Donald E.

2008-01-07

The Hanford Internal Dosimetry Program has adopted the computer code IMBA (Integrated Modules for Bioassay Analysis) as its primary code for bioassay data evaluation and dose assessment using methodologies of ICRP Publications 60, 66, 67, 68, and 78. The adoption of this code was part of the implementation plan for the June 8, 2007 amendments to 10 CFR 835. This information release includes action items unique to IMBA that were required by PNNL quality assurance standards for implementation of safety software. Copie of the IMBA software verification test plan and the outline of the briefing given to new users aremore » also included.« less

Formal Safety Certification of Aerospace Software

NASA Technical Reports Server (NTRS)

Denney, Ewen; Fischer, Bernd

2005-01-01

In principle, formal methods offer many advantages for aerospace software development: they can help to achieve ultra-high reliability, and they can be used to provide evidence of the reliability claims which can then be subjected to external scrutiny. However, despite years of research and many advances in the underlying formalisms of specification, semantics, and logic, formal methods are not much used in practice. In our opinion this is related to three major shortcomings. First, the application of formal methods is still expensive because they are labor- and knowledge-intensive. Second, they are difficult to scale up to complex systems because they are based on deep mathematical insights about the behavior of the systems (t.e., they rely on the "heroic proof"). Third, the proofs can be difficult to interpret, and typically stand in isolation from the original code. In this paper, we describe a tool for formally demonstrating safety-relevant aspects of aerospace software, which largely circumvents these problems. We focus on safely properties because it has been observed that safety violations such as out-of-bounds memory accesses or use of uninitialized variables constitute the majority of the errors found in the aerospace domain. In our approach, safety means that the program will not violate a set of rules that can range for the simple memory access rules to high-level flight rules. These different safety properties are formalized as different safety policies in Hoare logic, which are then used by a verification condition generator along with the code and logical annotations in order to derive formal safety conditions; these are then proven using an automated theorem prover. Our certification system is currently integrated into a model-based code generation toolset that generates the annotations together with the code. However, this automated formal certification technology is not exclusively constrained to our code generator and could, in principle, also be integrated with other code generators such as RealTime Workshop or even applied to legacy code. Our approach circumvents the historical problems with formal methods by increasing the degree of automation on all levels. The restriction to safety policies (as opposed to arbitrary functional behavior) results in simpler proof problems that can generally be solved by fully automatic theorem proves. An automated linking mechanism between the safety conditions and the code provides some of the traceability mandated by process standards such as DO-178B. An automated explanation mechanism uses semantic markup added by the verification condition generator to produce natural-language explanations of the safety conditions and thus supports their interpretation in relation to the code. It shows an automatically generated certification browser that lets users inspect the (generated) code along with the safety conditions (including textual explanations), and uses hyperlinks to automate tracing between the two levels. Here, the explanations reflect the logical structure of the safety obligation but the mechanism can in principle be customized using different sets of domain concepts. The interface also provides some limited control over the certification process itself. Our long-term goal is a seamless integration of certification, code generation, and manual coding that results in a "certified pipeline" in which specifications are automatically transformed into executable code, together with the supporting artifacts necessary for achieving and demonstrating the high level of assurance needed in the aerospace domain.

SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 1, Part 2: Control modules S1--H1; Revision 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system.« less

Thermal-hydraulic modeling needs for passive reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, J.M.

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered,more » but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.« less

Pressure Safety Program Implementation at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lower, Mark; Etheridge, Tom; Oland, C. Barry

2013-01-01

The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According tomore » 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply with pressure safety requirements in 10 CFR 851. It also describes actions taken to develop and implement ORNL’s Pressure Safety Program.« less

Patient complaints in healthcare systems: a systematic review and coding taxonomy

PubMed Central

Reader, Tom W; Gillespie, Alex; Roberts, Jane

2014-01-01

Background Patient complaints have been identified as a valuable resource for monitoring and improving patient safety. This article critically reviews the literature on patient complaints, and synthesises the research findings to develop a coding taxonomy for analysing patient complaints. Methods The PubMed, Science Direct and Medline databases were systematically investigated to identify patient complaint research studies. Publications were included if they reported primary quantitative data on the content of patient-initiated complaints. Data were extracted and synthesised on (1) basic study characteristics; (2) methodological details; and (3) the issues patients complained about. Results 59 studies, reporting 88 069 patient complaints, were included. Patient complaint coding methodologies varied considerably (eg, in attributing single or multiple causes to complaints). In total, 113 551 issues were found to underlie the patient complaints. These were analysed using 205 different analytical codes which when combined represented 29 subcategories of complaint issue. The most common issues complained about were ‘treatment’ (15.6%) and ‘communication’ (13.7%). To develop a patient complaint coding taxonomy, the subcategories were thematically grouped into seven categories, and then three conceptually distinct domains. The first domain related to complaints on the safety and quality of clinical care (representing 33.7% of complaint issues), the second to the management of healthcare organisations (35.1%) and the third to problems in healthcare staff–patient relationships (29.1%). Conclusions Rigorous analyses of patient complaints will help to identify problems in patient safety. To achieve this, it is necessary to standardise how patient complaints are analysed and interpreted. Through synthesising data from 59 patient complaint studies, we propose a coding taxonomy for supporting future research and practice in the analysis of patient complaint data. PMID:24876289

Methodology, status and plans for development and assessment of TUF and CATHENA codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luxat, J.C.; Liu, W.S.; Leung, R.K.

1997-07-01

An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically onmore » CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.« less

Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Sumner, Tyler S.

2016-04-17

An advanced system analysis tool SAM is being developed for fast-running, improved-fidelity, and whole-plant transient analyses at Argonne National Laboratory under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. As an important part of code development, companion validation activities are being conducted to ensure the performance and validity of the SAM code. This paper presents the benchmark simulations of two EBR-II tests, SHRT-45R and BOP-302R, whose data are available through the support of DOE-NE’s Advanced Reactor Technology (ART) program. The code predictions of major primary coolant system parameter are compared with the test results. Additionally, the SAS4A/SASSYS-1 code simulationmore » results are also included for a code-to-code comparison.« less

Coupling of TRAC-PF1/MOD2, Version 5.4.25, with NESTLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knepper, P.L.; Hochreiter, L.E.; Ivanov, K.N.

1999-09-01

A three-dimensional (3-D) spatial kinetics capability within a thermal-hydraulics system code provides a more correct description of the core physics during reactor transients that involve significant variations in the neutron flux distribution. Coupled codes provide the ability to forecast safety margins in a best-estimate manner. The behavior of a reactor core and the feedback to the plant dynamics can be accurately simulated. For each time step, coupled codes are capable of resolving system interaction effects on neutronics feedback and are capable of describing local neutronics effects caused by the thermal hydraulics and neutronics coupling. With the improvements in computational technology,more » modeling complex reactor behaviors with coupled thermal hydraulics and spatial kinetics is feasible. Previously, reactor analysis codes were limited to either a detailed thermal-hydraulics model with simplified kinetics or multidimensional neutron kinetics with a simplified thermal-hydraulics model. The authors discuss the coupling of the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, with the NESTLE code.« less

Stress Analysis of the External Stores DP Ejection Launcher/Inner Tube Subsystem.

DTIC Science & Technology

1978-11-01

for Stress and Strain.2 The materials used throughout this analysis are 304, 316, and 17 - 4Ph corrosion resistant steel, AlIS5700 series steel, and 6061...Analysis; Inner Tube Subassembly; Safety Factors; External Stores (Doc Des--P) 17 . GIDEP REPRESENTATIVE 11R PARTICIPANT ACTIVITY AND CODE M. H. Sloan Naval... 17 2.8 Bracket, Cylinder ......................................... 19 2.9 Mounting Screws, Bracket

LRFD software for design and actual ultimate capacity of confined rectangular columns : [technical summary].

DOT National Transportation Integrated Search

2013-04-01

Columns are considered the most critical elements in structures. The unconfined analysis for columns is well established in the literature. Structural design codes dictate reduction factors for safety. It wasnt until very recently that design spec...

29 CFR 1910.35 - Compliance with alternate exit-route codes.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 1910.35 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR OCCUPATIONAL SAFETY AND HEALTH STANDARDS Exit Routes and Emergency Planning § 1910.35...-route provisions of NFPA 101, Life Safety Code, 2009 edition, or the exit-route provisions of the...

29 CFR 1910.144 - Safety color code for marking physical hazards.

Code of Federal Regulations, 2011 CFR

2011-07-01

... the basic color for the identification of: (i) Fire protection equipment and apparatus. [Reserved] (ii... 29 Labor 5 2011-07-01 2011-07-01 false Safety color code for marking physical hazards. 1910.144 Section 1910.144 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH...

Parallel computation of multigroup reactivity coefficient using iterative method

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

Results of SEI Independent Research and Development Projects and Report on Emerging Technologies and Technology Trends

DTIC Science & Technology

2004-10-01

Top-Level Process for Identification and Analysis of Safety-Related Re- quirements 4.4 Collaborators The primary SEI team members were Don Firesmith...Graff, M. & van Wyk, K. Secure Coding Principles & Practices. O’Reilly, 2003. • Hoglund, G. & McGraw, G. Exploiting Software: How to Break Code. Addison...Eisenecker, U.; Glück, R.; Vandevoorde, D.; & Veldhuizen , T. “Generative Programming and Active Libraries (Extended Abstract)” <osl.iu.edu/~tveldhui/papers

Structural mechanics simulations

NASA Technical Reports Server (NTRS)

Biffle, Johnny H.

1992-01-01

Sandia National Laboratory has a very broad structural capability. Work has been performed in support of reentry vehicles, nuclear reactor safety, weapons systems and components, nuclear waste transport, strategic petroleum reserve, nuclear waste storage, wind and solar energy, drilling technology, and submarine programs. The analysis environment contains both commercial and internally developed software. Included are mesh generation capabilities, structural simulation codes, and visual codes for examining simulation results. To effectively simulate a wide variety of physical phenomena, a large number of constitutive models have been developed.

Safety and health in the construction of fixed offshore installations in the petroleum industry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-01-01

A meeting convened by the ILO (International Labor Office) on safety problems in the offshore petroleum industry recommended the preparation of a code of practice setting out standards for safety and health during the construction of fixed offshore installations. Such a code, to be prepared by the ILO in co-operation with other bodies, including the Inter-Governmental Maritime Consultative Organisation (IMCO), was to take into consideration existing standards applicable to offshore construction activities and to supplement the ILO codes of practice on safety and health in building and civil engineering work, shipbuilding and ship repairing. (Copyright (c) International Labour Organisation 1981.)

Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

DOE PAGES

Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; ...

2014-11-01

This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

42 CFR 482.41 - Condition of participation: Physical environment.

Code of Federal Regulations, 2013 CFR

2013-10-01

...: Life safety from fire. (1) Except as otherwise provided in this section— (i) The hospital must meet the applicable provisions of the 2000 edition of the Life Safety Code of the National Fire Protection Association... Life Safety Code, issued January 14, 2000, for incorporation by reference in accordance with 5 U.S.C...

42 CFR 482.41 - Condition of participation: Physical environment.

Code of Federal Regulations, 2014 CFR

2014-10-01

...: Life safety from fire. (1) Except as otherwise provided in this section— (i) The hospital must meet the applicable provisions of the 2000 edition of the Life Safety Code of the National Fire Protection Association... Life Safety Code, issued January 14, 2000, for incorporation by reference in accordance with 5 U.S.C...

Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

DOE Office of Scientific and Technical Information (OSTI.GOV)

First, M.W.

1991-02-01

Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

Review of current nuclear fallout codes.

PubMed

Auxier, Jerrad P; Auxier, John D; Hall, Howard L

2017-05-01

The importance of developing a robust nuclear forensics program to combat the illicit use of nuclear material that may be used as an improvised nuclear device is widely accepted. In order to decrease the threat to public safety and improve governmental response, government agencies have developed fallout-analysis codes to predict the fallout particle size, dose, and dispersion and dispersion following a detonation. This paper will review the different codes that have been developed for predicting fallout from both chemical and nuclear weapons. This will decrease the response time required for the government to respond to the event. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

The U. S. Department of Energy SARP review training program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mauck, C.J.

1988-01-01

In support of its radioactive material packaging certification program, the U.S. Department of Energy (DOE) has established a special training workshop. The purpose of the two-week workshop is to develop skills in reviewing Safety Analysis Reports for Packagings (SARPs) and performing confirmatory analyses. The workshop, conducted by the Lawrence Livermore National Laboratory (LLNL) for DOE, is divided into two parts: methods of review and methods of analysis. The sessions covering methods of review are based on the DOE document, ''Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings'' (PRG). The sessions cover relevant DOE Orders and all areas ofmore » review in the applicable Nuclear Regulatory Commission (NRC) Regulatory Guides. The technical areas addressed include structural and thermal behavior, materials, shielding, criticality, and containment. The course sessions on methods of analysis provide hands-on experience in the use of calculational methods and codes for reviewing SARPs. Analytical techniques and computer codes are discussed and sample problems are worked. Homework is assigned each night and over the included weekend; at the conclusion, a comprehensive take-home examination is given requiring six to ten hours to complete.« less

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Energy Storage System Safety: Plan Review and Inspection Checklist

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cole, Pam C.; Conover, David R.

Codes, standards, and regulations (CSR) governing the design, construction, installation, commissioning, and operation of the built environment are intended to protect the public health, safety, and welfare. While these documents change over time to address new technology and new safety challenges, there is generally some lag time between the introduction of a technology into the market and the time it is specifically covered in model codes and standards developed in the voluntary sector. After their development, there is also a timeframe of at least a year or two until the codes and standards are adopted. Until existing model codes andmore » standards are updated or new ones are developed and then adopted, one seeking to deploy energy storage technologies or needing to verify the safety of an installation may be challenged in trying to apply currently implemented CSRs to an energy storage system (ESS). The Energy Storage System Guide for Compliance with Safety Codes and Standards1 (CG), developed in June 2016, is intended to help address the acceptability of the design and construction of stationary ESSs, their component parts, and the siting, installation, commissioning, operations, maintenance, and repair/renovation of ESS within the built environment.« less

Report on SARS backfit evaluation, Catalytic, Inc. Solvent Refined Coal Pilot Plant, Wilsonville, Alabama

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meyer, A.F. Jr.

1980-07-02

A site visit was made in company with the DOE-OPTA-EA Safety and Health Official for the purpose of providing that official with technical assistance in evaluating the validity of an earlier DOE-OPTA recommendation exempting this facility from the Safety and Analysis and Review backfit requirements of DOE Order 5481.1. A further purpose of the visit was to assess and evaluate the occupational safety and health program at this facility, as compared with the criteria and guidelines contained in ASFE Order 5481.1. Adequate documentation regarding compliance with codes, standards, and regulations were observed at this facility. There is in existence anmore » ongoing continuous safety analysis effort for both modifications or additions to this facility. Adequate environmental safeguards and plans and procedures were observed. The SARS backfit exemption is appropriate. The occupational safety and health program is in many ways a model for the scope of work and nature of hazards involved, and is consistent with ASFE guidelines and statutory requirements.« less

2009 Annual Progress Report: DOE Hydrogen Program, November 2009 (Book)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

2009-11-01

This report summarizes the hydrogen and fuel cell R&D activities and accomplishments of the DOE Hydrogen Program for FY2009. It covers the program areas of hydrogen production and delivery; fuel cells; manufacturing; technology validation; safety, codes and standards; education; and systems analysis.

Medicare and Medicaid Programs; Fire Safety Requirements for Certain Health Care Facilities. Final rule.

PubMed

2016-05-04

This final rule will amend the fire safety standards for Medicare and Medicaid participating hospitals, critical access hospitals (CAHs), long-term care facilities, intermediate care facilities for individuals with intellectual disabilities (ICF-IID), ambulatory surgery centers (ASCs), hospices which provide inpatient services, religious non-medical health care institutions (RNHCIs), and programs of all-inclusive care for the elderly (PACE) facilities. Further, this final rule will adopt the 2012 edition of the Life Safety Code (LSC) and eliminate references in our regulations to all earlier editions of the Life Safety Code. It will also adopt the 2012 edition of the Health Care Facilities Code, with some exceptions.

Analyses in support of risk-informed natural gas vehicle maintenance facility codes and standards :

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ekoto, Isaac W.; Blaylock, Myra L.; LaFleur, Angela Christine

2014-03-01

Safety standards development for maintenance facilities of liquid and compressed gas fueled large-scale vehicles is required to ensure proper facility design and operation envelopes. Standard development organizations are utilizing risk-informed concepts to develop natural gas vehicle (NGV) codes and standards so that maintenance facilities meet acceptable risk levels. The present report summarizes Phase I work for existing NGV repair facility code requirements and highlights inconsistencies that need quantitative analysis into their effectiveness. A Hazardous and Operability study was performed to identify key scenarios of interest. Finally, scenario analyses were performed using detailed simulations and modeling to estimate the overpressure hazardsmore » from HAZOP defined scenarios. The results from Phase I will be used to identify significant risk contributors at NGV maintenance facilities, and are expected to form the basis for follow-on quantitative risk analysis work to address specific code requirements and identify effective accident prevention and mitigation strategies.« less

2014 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

2014-11-01

The 2014 Annual Progress Report summarizes fiscal year 2014 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

This report summarizes comments from the Peer Review Panel at the 2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 13-17, 2013, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

Evaluation of liquefaction potential for building code

NASA Astrophysics Data System (ADS)

Nunziata, C.; De Nisco, G.; Panza, G. F.

2008-07-01

The standard approach for the evaluation of the liquefaction susceptibility is based on the estimation of a safety factor between the cyclic shear resistance to liquefaction and the earthquake induced shear stress. Recently, an updated procedure based on shear-wave velocities (Vs) has been proposed which could be more easily applied. These methods have been applied at La Plaja beach of Catania, that experienced liquefaction because of the 1693 earthquake. The detailed geotechnical and Vs information and the realistic ground motion computed for the 1693 event let us compare the two approaches. The successful application of the Vs procedure, slightly modified to fit historical and safety factor information, even if additional field performances are needed, encourages the development of a guide for liquefaction potential analysis, based on well defined Vs profiles to be included in the italian seismic code.

Common Day Care Safety Renovations: Descriptions, Explanations and Cost Estimates.

ERIC Educational Resources Information Center

Spack, Stan

This booklet explains some of the day care safety features specified by the new Massachusetts State Building Code (January 1, 1975) which must be met before a new day care center can be licensed. The safety features described are those which most often require renovation to meet the building code standards. Best estimates of the costs involved in…

Image processing for safety assessment in civil engineering.

PubMed

Ferrer, Belen; Pomares, Juan C; Irles, Ramon; Espinosa, Julian; Mas, David

2013-06-20

Behavior analysis of construction safety systems is of fundamental importance to avoid accidental injuries. Traditionally, measurements of dynamic actions in civil engineering have been done through accelerometers, but high-speed cameras and image processing techniques can play an important role in this area. Here, we propose using morphological image filtering and Hough transform on high-speed video sequence as tools for dynamic measurements on that field. The presented method is applied to obtain the trajectory and acceleration of a cylindrical ballast falling from a building and trapped by a thread net. Results show that safety recommendations given in construction codes can be potentially dangerous for workers.

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Malley, Kathleen; Lopez, Hugo; Cairns, Julie

An overview of the main North American codes and standards associated with hydrogen safety sensors is provided. The distinction between a code and a standard is defined, and the relationship between standards and codes is clarified, especially for those circumstances where a standard or a certification requirement is explicitly referenced within a code. The report identifies three main types of standards commonly applied to hydrogen sensors (interface and controls standards, shock and hazard standards, and performance-based standards). The certification process and a list and description of the main standards and model codes associated with the use of hydrogen safety sensorsmore » in hydrogen infrastructure are presented.« less

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

NASA Astrophysics Data System (ADS)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Annotation analysis for testing drug safety signals using unstructured clinical notes

PubMed Central

2012-01-01

Background The electronic surveillance for adverse drug events is largely based upon the analysis of coded data from reporting systems. Yet, the vast majority of electronic health data lies embedded within the free text of clinical notes and is not gathered into centralized repositories. With the increasing access to large volumes of electronic medical data—in particular the clinical notes—it may be possible to computationally encode and to test drug safety signals in an active manner. Results We describe the application of simple annotation tools on clinical text and the mining of the resulting annotations to compute the risk of getting a myocardial infarction for patients with rheumatoid arthritis that take Vioxx. Our analysis clearly reveals elevated risks for myocardial infarction in rheumatoid arthritis patients taking Vioxx (odds ratio 2.06) before 2005. Conclusions Our results show that it is possible to apply annotation analysis methods for testing hypotheses about drug safety using electronic medical records. PMID:22541596

Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merrill, Brad J

This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allowsmore » the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.« less

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

Code development for ships -- A demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayyub, B.; Mansour, A.E.; White, G.

1996-12-31

A demonstration summary of a reliability-based structural design code for ships is presented for two ship types, a cruiser and a tanker. For both ship types, code requirements cover four failure modes: hull girder bulking, unstiffened plate yielding and buckling, stiffened plate buckling, and fatigue of critical detail. Both serviceability and ultimate limit states are considered. Because of limitation on the length, only hull girder modes are presented in this paper. Code requirements for other modes will be presented in future publication. A specific provision of the code will be a safety check expression. The design variables are to bemore » taken at their nominal values, typically values in the safe side of the respective distributions. Other safety check expressions for hull girder failure that include load combination factors, as well as consequence of failure factors, are considered. This paper provides a summary of safety check expressions for the hull girder modes.« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.

2011-03-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repositorymore » designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Curtis; Mandelli, Diego; Prescott, Steven

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools.more » This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.« less

Using Modified-ISS Model to Evaluate Medication Administration Safety During Bar Code Medication Administration Implementation in Taiwan Regional Teaching Hospital.

PubMed

Ma, Pei-Luen; Jheng, Yan-Wun; Jheng, Bi-Wei; Hou, I-Ching

2017-01-01

Bar code medication administration (BCMA) could reduce medical errors and promote patient safety. This research uses modified information systems success model (M-ISS model) to evaluate nurses' acceptance to BCMA. The result showed moderate correlation between medication administration safety (MAS) to system quality, information quality, service quality, user satisfaction, and limited satisfaction.

MODEL 9977 B(M)F-96 SAFETY ANALYSIS REPORT FOR PACKAGING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abramczyk, G; Paul Blanton, P; Kurt Eberl, K

2006-05-18

This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on and for the 9977 Shipping Package, referred to as the General Purpose Fissile Package (GPFP). The performance evaluation presented in this SARP documents the compliance of the 9977 package with the regulatory safety requirements for Type B packages. Per 10 CFR 71.59, for the 9977 packages evaluated in this SARP, the value of ''N'' is 50, and the Transport Index based on nuclear criticality control is 1.0. The 9977 package is designed with a high degree of single containment. The 9977 complies with 10 CFR 71more » (2002), Department of Energy (DOE) Order 460.1B, DOE Order 460.2, and 10 CFR 20 (2003) for As Low As Reasonably Achievable (ALARA) principles. The 9977 also satisfies the requirements of the Regulations for the Safe Transport of Radioactive Material--1996 Edition (Revised)--Requirements. IAEA Safety Standards, Safety Series No. TS-R-1 (ST-1, Rev.), International Atomic Energy Agency, Vienna, Austria (2000). The 9977 package is designed, analyzed and fabricated in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1992 edition.« less

RAVEN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph

2017-03-01

RAVEN is a generic software framework to perform parametric and probabilistic analy- sis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncer- tainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters thatmore » need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused to- ward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is in- terested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an important process such as peak pressure in a pipe) is exceeded under certain conditions. The initial development of RAVEN has been focused on providing dynamic risk assess- ment capability to RELAP-7, currently under develop-ment at the INL and, likely, future replacement of the RELAP5-3D code. Most the capabilities that have been implemented having RELAP-7 as principal focus are easily deployable for other system codes. For this reason, several side activates are currently ongoing for coupling RAVEN with soft- ware such as RELAP5-3D, etc. The aim of this document is the explaination of the input requirements, focalizing on the input structure.« less

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Promoting personal safety of building service workers: issues and challenges.

PubMed

Chen, Shelley I; Skillen, D Lynn

2006-06-01

This exploratory, descriptive study conducted at a large western Canadian university solicited perceptions of personal safety among building service workers who perform night shift work alone. Ten semi-structured interviews were conducted at approximately 10:00 p.m. or 7:00 a.m with a convenience sample of night building service workers in private or semi-private locations on the university campus. Transcribed interview data were subjected to inductive content analysis using descriptive, interpretive, and pattern coding (Miles & Huberman, 1994). Results suggest that building service night shift workers are exposed to personal safety hazards in their physical and psychosocial work environments. In addition, culturally and linguistically appropriate delivery of safety training and education about policies and procedures is required for culturally diverse building service workers. Promotion of personal safety in this heterogeneous worker population requires due diligence, assessment, and advocacy.

Scale/TSUNAMI Sensitivity Data for ICSBEP Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T; Reed, Davis Allan; Lefebvre, Robert A

2011-01-01

The Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) software developed at Oak Ridge National Laboratory (ORNL) as part of the Scale code system provide unique methods for code validation, gap analysis, and experiment design. For TSUNAMI analysis, sensitivity data are generated for each application and each existing or proposed experiment used in the assessment. The validation of diverse sets of applications requires potentially thousands of data files to be maintained and organized by the user, and a growing number of these files are available through the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) distributed through themore » International Criticality Safety Benchmark Evaluation Program (ICSBEP). To facilitate the use of the IHECSBE benchmarks in rigorous TSUNAMI validation and gap analysis techniques, ORNL generated SCALE/TSUNAMI sensitivity data files (SDFs) for several hundred benchmarks for distribution with the IHECSBE. For the 2010 edition of IHECSBE, the sensitivity data were generated using 238-group cross-section data based on ENDF/B-VII.0 for 494 benchmark experiments. Additionally, ORNL has developed a quality assurance procedure to guide the generation of Scale inputs and sensitivity data, as well as a graphical user interface to facilitate the use of sensitivity data in identifying experiments and applying them in validation studies.« less

Practical guide to bar coding for patient medication safety.

PubMed

Neuenschwander, Mark; Cohen, Michael R; Vaida, Allen J; Patchett, Jeffrey A; Kelly, Jamie; Trohimovich, Barbara

2003-04-15

Bar coding for the medication administration step of the drug-use process is discussed. FDA will propose a rule in 2003 that would require bar-code labels on all human drugs and biologicals. Even with an FDA mandate, manufacturer procrastination and possible shifts in product availability are likely to slow progress. Such delays should not preclude health systems from adopting bar-code-enabled point-of-care (BPOC) systems to achieve gains in patient safety. Bar-code technology is a replacement for traditional keyboard data entry. The elements of bar coding are content, which determines the meaning; data format, which refers to the embedded data and symbology, which describes the "font" in which the machine-readable code is written. For a BPOC system to deliver an acceptable level of patient protection, the hospital must first establish reliable processes for a patient identification band, caregiver badge, and medication bar coding. Medications can have either drug-specific or patient-specific bar codes. Both varieties result in the desired code that supports patient's five rights of drug administration. When medications are not available from the manufacturer in immediate-container bar-coded packaging, other means of applying the bar code must be devised, including the use of repackaging equipment, overwrapping, manual bar coding, and outsourcing. Virtually all medications should be bar coded, the bar code on the label should be easily readable, and appropriate policies, procedures, and checks should be in place. Bar coding has the potential to be not only cost-effective but to produce a return on investment. By bar coding patient identification tags, caregiver badges, and immediate-container medications, health systems can substantially increase patient safety during medication administration.

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

2008 DOE Hydrogen Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

2008-06-13

This report summarizes comments from the Peer Review Panel at the 2008 DOE Hydrogen Program Annual Merit Review, held on June 9-13, 2008, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; technology validation; safety, codes, and standards; education; systems analysis; and manufacturing.

2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The 2015 Annual Progress Report summarizes fiscal year 2015 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; and market transformation.

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

78 FR 23497 - Propiconazole; Pesticide Tolerances

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-19

...). Animal production (NAICS code 112). Food manufacturing (NAICS code 311). Pesticide manufacturing (NAICS.... Aggregate Risk Assessment and Determination of Safety Section 408(b)(2)(A)(i) of FFDCA allows EPA to... dose at which adverse effects of concern are identified (the LOAEL). Uncertainty/safety factors are...

Code of Sustainable Practice in Occupational and Environmental Health and Safety for Corporations.

PubMed

Castleman, Barry; Allen, Barbara; Barca, Stefania; Bohme, Susanna Rankin; Henry, Emmanuel; Kaur, Amarjit; Massard-Guilbaud, Genvieve; Melling, Joseph; Menendez-Navarro, Alfredo; Renfrew, Daniel; Santiago, Myrna; Sellers, Christopher; Tweedale, Geoffrey; Zalik, Anna; Zavestoski, Stephen

2008-01-01

At a conference held at Stony Brook University in December 2007, "Dangerous Trade: Histories of Industrial Hazard across a Globalizing World," participants endorsed a Code of Sustainable Practice in Occupational and Environmental Health and Safety for Corporations. The Code outlines practices that would ensure corporations enact the highest health and environmentally protective measures in all the locations in which they operate. Corporations should observe international guidelines on occupational exposure to air contaminants, plant safety, air and water pollutant releases, hazardous waste disposal practices, remediation of polluted sites, public disclosure of toxic releases, product hazard labeling, sale of products for specific uses, storage and transport of toxic intermediates and products, corporate safety and health auditing, and corporate environmental auditing. Protective measures in all locations should be consonant with the most protective measures applied anywhere in the world, and should apply to the corporations' subsidiaries, contractors, suppliers, distributors, and licensees of technology. Key words: corporations, sustainability, environmental protection, occupational health, code of practice.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Patient safety awareness among Undergraduate Medical Students in Pakistani Medical School.

PubMed

Kamran, Rizwana; Bari, Attia; Khan, Rehan Ahmed; Al-Eraky, Mohamed

2018-01-01

To measure the level of awareness of patient safety among undergraduate medical students in Pakistani Medical School and to find the difference with respect to gender and prior experience with medical error. This cross-sectional study was conducted at the University of Lahore (UOL), Pakistan from January to March 2017, and comprised final year medical students. Data was collected using a questionnaire 'APSQ- III' on 7 point Likert scale. Eight questions were reverse coded. Survey was anonymous. SPSS package 20 was used for statistical analysis. Questionnaire was filled by 122 students, with 81% response rate. The best score 6.17 was given for the 'team functioning', followed by 6.04 for 'long working hours as a cause of medical error'. The domains regarding involvement of patient, confidence to report medical errors and role of training and learning on patient safety scored high in the agreed range of >5. Reverse coded questions about 'professional incompetence as an error cause' and 'disclosure of errors' showed negative perception. No significant differences of perceptions were found with respect to gender and prior experience with medical error (p= >0.05). Undergraduate medical students at UOL had a positive attitude towards patient safety. However, there were misconceptions about causes of medical errors and error disclosure among students and patient safety education needs to be incorporated in medical curriculum of Pakistan.

Patient safety awareness among Undergraduate Medical Students in Pakistani Medical School

PubMed Central

Kamran, Rizwana; Bari, Attia; Khan, Rehan Ahmed; Al-Eraky, Mohamed

2018-01-01

Objective: To measure the level of awareness of patient safety among undergraduate medical students in Pakistani Medical School and to find the difference with respect to gender and prior experience with medical error. Methods: This cross-sectional study was conducted at the University of Lahore (UOL), Pakistan from January to March 2017, and comprised final year medical students. Data was collected using a questionnaire ‘APSQ- III’ on 7 point Likert scale. Eight questions were reverse coded. Survey was anonymous. SPSS package 20 was used for statistical analysis. Results: Questionnaire was filled by 122 students, with 81% response rate. The best score 6.17 was given for the ‘team functioning’, followed by 6.04 for ‘long working hours as a cause of medical error’. The domains regarding involvement of patient, confidence to report medical errors and role of training and learning on patient safety scored high in the agreed range of >5. Reverse coded questions about ‘professional incompetence as an error cause’ and ‘disclosure of errors’ showed negative perception. No significant differences of perceptions were found with respect to gender and prior experience with medical error (p= >0.05). Conclusion: Undergraduate medical students at UOL had a positive attitude towards patient safety. However, there were misconceptions about causes of medical errors and error disclosure among students and patient safety education needs to be incorporated in medical curriculum of Pakistan. PMID:29805398

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

Safety and Liability.

ERIC Educational Resources Information Center

Berthelot, Ronald J.; And Others

1982-01-01

This series of five articles highlights Pensacola Junior College's occupational safety course, involving simulated emergencies, Florida's standards for teacher liability, electrical safety in the classroom and laboratory, color coding for machine safety, and Florida industrial arts safety instructional materials. (SK)

SU-F-P-10: A Web-Based Radiation Safety Relational Database Module for Regulatory Compliance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosen, C; Ramsay, B; Konerth, S

Purpose: Maintaining compliance with Radioactive Materials Licenses is inherently a time-consuming task requiring focus and attention to detail. Staff tasked with these responsibilities, such as the Radiation Safety Officer and associated personnel must retain disparate records for eventual placement into one or more annual reports. Entering results and records in a relational database using a web browser as the interface, and storing that data in a cloud-based storage site, removes procedural barriers. The data becomes more adaptable for mining and sharing. Methods: Web-based code was written utilizing the web framework Django, written in Python. Additionally, the application utilizes JavaScript formore » front-end interaction, SQL, HTML and CSS. Quality assurance code testing is performed in a sequential style, and new code is only added after the successful testing of the previous goals. Separate sections of the module include data entry and analysis for audits, surveys, quality management, and continuous quality improvement. Data elements can be adapted for quarterly and annual reporting, and for immediate notification of user determined alarm settings. Results: Current advances are focusing on user interface issues, and determining the simplest manner by which to teach the user to build query forms. One solution has been to prepare library documents that a user can select or edit in place of creation a new document. Forms are being developed based upon Nuclear Regulatory Commission federal code, and will be expanded to include State Regulations. Conclusion: Establishing a secure website to act as the portal for data entry, storage and manipulation can lead to added efficiencies for a Radiation Safety Program. Access to multiple databases can lead to mining for big data programs, and for determining safety issues before they occur. Overcoming web programming challenges, a category that includes mathematical handling, is providing challenges that are being overcome.« less

RETRANO3 benchmarks for Beaver Valley plant transients and FSAR analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beaumont, E.T.; Feltus, M.A.

1993-01-01

Any best-estimate code (e.g., RETRANO3) results must be validated against plant data and final safety analysis report (FSAR) predictions. The need for two independent means of benchmarking is necessary to ensure that the results were not biased toward a particular data set and to have a certain degree of accuracy. The code results need to be compared with previous results and show improvements over previous code results. Ideally, the two best means of benchmarking a thermal hydraulics code are comparing results from previous versions of the same code along with actual plant data. This paper describes RETRAN03 benchmarks against RETRAN02more » results, actual plant data, and FSAR predictions. RETRAN03, the Electric Power Research Institute's latest version of the RETRAN thermal-hydraulic analysis codes, offers several upgrades over its predecessor, RETRAN02 Mod5. RETRAN03 can use either implicit or semi-implicit numerics, whereas RETRAN02 Mod5 uses only semi-implicit numerics. Another major upgrade deals with slip model options. RETRAN03 added several new models, including a five-equation model for more accurate modeling of two-phase flow. RETPAN02 Mod5 should give similar but slightly more conservative results than RETRAN03 when executed with RETRAN02 Mod5 options.« less

Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies willmore » still be present in the successor code RELAP5/MOD3.« less

Guidelines for the Design, Fabrication, Testing, Installation and Operation of Srf Cavities

NASA Astrophysics Data System (ADS)

Theilacker, J.; Carter, H.; Foley, M.; Hurh, P.; Klebaner, A.; Krempetz, K.; Nicol, T.; Olis, D.; Page, T.; Peterson, T.; Pfund, P.; Pushka, D.; Schmitt, R.; Wands, R.

2010-04-01

Superconducting Radio-Frequency (SRF) cavities containing cryogens under pressure pose a potential rupture hazard to equipment and personnel. Generally, pressure vessels fall within the scope of the ASME Boiler and Pressure Vessel Code however, the use of niobium as a material for the SRF cavities is beyond the applicability of the Code. Fermilab developed a guideline to ensure sound engineering practices governing the design, fabrication, testing, installation and operation of SRF cavities. The objective of the guideline is to reduce hazards and to achieve an equivalent level of safety afforded by the ASME Code. The guideline addresses concerns specific to SRF cavities in the areas of materials, design and analysis, welding and brazing, pressure relieving requirements, pressure testing and quality control.

Synthesizing Certified Code

NASA Technical Reports Server (NTRS)

Whalen, Michael; Schumann, Johann; Fischer, Bernd

2002-01-01

Code certification is a lightweight approach to demonstrate software quality on a formal level. Its basic idea is to require producers to provide formal proofs that their code satisfies certain quality properties. These proofs serve as certificates which can be checked independently. Since code certification uses the same underlying technology as program verification, it also requires many detailed annotations (e.g., loop invariants) to make the proofs possible. However, manually adding theses annotations to the code is time-consuming and error-prone. We address this problem by combining code certification with automatic program synthesis. We propose an approach to generate simultaneously, from a high-level specification, code and all annotations required to certify generated code. Here, we describe a certification extension of AUTOBAYES, a synthesis tool which automatically generates complex data analysis programs from compact specifications. AUTOBAYES contains sufficient high-level domain knowledge to generate detailed annotations. This allows us to use a general-purpose verification condition generator to produce a set of proof obligations in first-order logic. The obligations are then discharged using the automated theorem E-SETHEO. We demonstrate our approach by certifying operator safety for a generated iterative data classification program without manual annotation of the code.

[Medication error management climate and perception for system use according to construction of medication error prevention system].

PubMed

Kim, Myoung Soo

2012-08-01

The purpose of this cross-sectional study was to examine current status of IT-based medication error prevention system construction and the relationships among system construction, medication error management climate and perception for system use. The participants were 124 patient safety chief managers working for 124 hospitals with over 300 beds in Korea. The characteristics of the participants, construction status and perception of systems (electric pharmacopoeia, electric drug dosage calculation system, computer-based patient safety reporting and bar-code system) and medication error management climate were measured in this study. The data were collected between June and August 2011. Descriptive statistics, partial Pearson correlation and MANCOVA were used for data analysis. Electric pharmacopoeia were constructed in 67.7% of participating hospitals, computer-based patient safety reporting systems were constructed in 50.8%, electric drug dosage calculation systems were in use in 32.3%. Bar-code systems showed up the lowest construction rate at 16.1% of Korean hospitals. Higher rates of construction of IT-based medication error prevention systems resulted in greater safety and a more positive error management climate prevailed. The supportive strategies for improving perception for use of IT-based systems would add to system construction, and positive error management climate would be more easily promoted.

Maternal worries, home safety behaviors, and perceived difficulties.

PubMed

Hendrickson, Sherry Garrett

2008-01-01

The aim of the study was to explore the worries, safety behaviors, and perceived difficulties in keeping children safe at home in a purposive sample of low-income, predominantly non-English speaking mothers as a foundation for later nursing interventions. This study was a qualitative, descriptive design with content analysis to identify maternal concerns, behaviors, and perceptions of home safety as part of a larger study. Eighty-two mothers, 64% of whom were monolingual Spanish-speakers, responded in writing to three semistructured interview questions. When mothers were unable to read and write the researcher wrote the responses, then read the content aloud for verification. A standardized probe for each question was posed to obtain richer responses. Data management included use of the software program NUD*IST and coding analyses following the Miles and Huberman guidelines (1994). Interpretations were translated into English for this report. The major worries were falling, health, kidnapping, and being hit by a car. The leading maternal behaviors were coded as being physically, verbally, and environmentally preventive. Mothers said that it was their role to provide safety, and that this role could be wearisome, such that constant supervision was difficult. Low-income mothers described their worries for their 1 to 4 year-old children, explored their behaviors for preventing injury, and discussed what made keeping children from harm difficult. Understanding how mothers keep children safe, the barriers to home safety, and effective safety behaviors are important to the health of children. The clinical relevance of this study includes building trust as clinicians plan assessment, intervention and evaluation of home safety to encourage dialog about concerns, safety behaviors, and barriers to keeping children from injury.

Methodology, status and plans for development and assessment of Cathare code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestion, D.; Barre, F.; Faydide, B.

1997-07-01

This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests ormore » integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.« less

Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

DOE PAGES

Williams, M. L.; Wiarda, D.; Ilas, G.; ...

2014-06-15

Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

Ocean Thermal Energy Conversion power system development. Phase I: preliminary design. Final report. [ODSP-3 code; OTEC Steady-State Analysis Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-12-04

The following appendices are included; Dynamic Simulation Program (ODSP-3); sample results of dynamic simulation; trip report - NH/sub 3/ safety precautions/accident records; trip report - US Coast Guard Headquarters; OTEC power system development, preliminary design test program report; medium turbine generator inspection point program; net energy analysis; bus bar cost of electricity; OTEC technical specifications; and engineer drawings. (WHK)

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, N.M.; Petrie, L.M.; Westfall, R.M.

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries.« less

Analysis of typical WWER-1000 severe accident scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sorokin, Yu.S.; Shchekoldin, V.V.; Borisov, L.N.

2004-07-01

At present in EDO 'Gidropress' there is a certain experience of performing the analyses of severe accidents of reactor plant with WWER with application of domestic and foreign codes. Important data were also obtained by the results of calculation modeling of integrated experiments with fuel assembly melting comprising a real fuel. Systematization and consideration of these data in development and assimilation of codes are extremely important in connection with large uncertainty still existing in understanding and adequate description of phenomenology of severe accidents. The presented report gives a comparison of analysis results of severe accidents of reactor plant with WWER-1000more » for two typical scenarios made by using American MELCOR code and the Russian RATEG/SVECHA/HEFEST code. The results of calculation modeling are compared using above codes with the data of experiment FPT1 with fuel assembly melting comprising a real fuel, which has been carried out at the facility Phebus (France). The obtained results are considered in the report from the viewpoint of: - adequacy of results of calculation modeling of separate phenomena during severe accidents of RP with WWER by using the above codes; - influence of uncertainties (degree of details of calculation models, choice of parameters of models etc.); - choice of those or other setup variables (options) in the used codes; - necessity of detailed modeling of processes and phenomena as applied to design justification of safety of RP with WWER. (authors)« less

Road safety practices among commercial motorcyclists in a rural town in Nigeria: implications for health education.

PubMed

Amoran, O E; Eme, Owoaje; Giwa, O A; Gbolahan, O B

This cross-sectional, community-based study was carried out among commercial motorcyclists in Igboora. All the commercial motor parks in Igboora were visited and all the commercial motorcyclists who consented to participate in the study were interviewed. Information on the respondents' socio-demographic characteristics, and the practice of road safety measures was collected using an interviewer administered questionnaire. A total of 299 motorcyclists were interviewed. All (100%) of them were males. The mean age of the respondents was 27.4 +/- 7.4 years. One hundred eighty-two (60.7%) of the motorcyclists had the correct knowledge of the purpose of Highway Code. Only 70 (23.3%) could recognize more than half of the currently used road safety codes and 47 (15.7%) obey these road safety codes more than half of the time they see it. Only 183 (61.2%) of them had a driving license and 72 (24.1%) were able to produce these licenses on demand. All (100%) of the respondents did not use any protective helmet. Those who have longer years of working experience, higher level of education and higher knowledge of the safety codes practice it more regularly (r = 0.198, p = 0.001, chi2= 9.31, p = 0.025, and r = 0.28, p = 0.001 respectively). One hundred thirty-six (45.5%) have been involved in at least one accident in the preceding year. The overall incidence of road traffic accident was 2.16 per 1,000. There was however on statistically significant association between the practice of road safety codes and the occurrence of road traffic accidents (chi2= 0.176, p = 0.916). The study shows that the practice of road safety measures was low in this rural Nigerian community and was not associated with the incidence of road traffic accidents. Introducing road safety education particularly targeted at educating the motorcyclists on the importance and practice of road safety measures would lead to an increase in the practice of the safety measures and hopefully a reduction in the incidence of road traffic accidents.

Measuring patient safety culture in maternal and child health institutions in China: a qualitative study.

PubMed

Wang, Yuanyuan; Liu, Weiwei; Shi, Huifeng; Liu, Chaojie; Wang, Yan

2017-07-12

Patient safety culture (PSC) plays a critical role in ensuring safe and quality care. Extensive PSC studies have been undertaken in hospitals. However, little is known about PSC in maternal and child health (MCH) institutions in China, which provide both population-based preventive services as well as individual care for patients. This study aimed to develop a theoretical framework for conceptualising PSC in MCH institutions in China. The study was undertaken in six MCH institutions (three in Hebei and three in Beijing). Participants (n=118) were recruited through stratified purposive sampling: 20 managers/administrators, 59 care providers and 39 patients. In-depth interviews were conducted with the participants. The interview data were coded using both inductive (based on the existing PSC theory developed by the Agency for Healthcare Research and Quality) and deductive (open coding arising from data) approaches. A PSC framework was formulated through axial coding that connected initial codes and selective coding that extracted a small number of themes. The interviewees considered patient safety in relation to six aspects: safety and security in public spaces, safety of medical services, privacy and information security, financial security, psychological safety and gap in services. A 12-dimensional PSC framework was developed, containing 69 items. While the existing PSC theory was confirmed by this study, some new themes emerged from the data. Patients expressed particular concerns about psychological safety and financial security. Defensive medical practices emerged as a PSC dimension that is associated with not only medical safety but also financial security and psychological safety. Patient engagement was also valued by the interviewees, especially the patients, as part of PSC. Although there are some common features in PSC across different healthcare delivery systems, PSC can also be context specific. In MCH settings in China, the meaning of 'patient safety' goes beyond the traditional definition of patients. General well-being, health and disease prevention are important anchor points for defining PSC in such settings. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

Application of the Life Safety Code to a Historic Test Stand

NASA Technical Reports Server (NTRS)

Askins, Bruce; Lemke, Paul R.; Lewis, William L.; Covell, Carol C.

2011-01-01

NASA has conducted a study to assess alternatives to refurbishing existing launch vehicle modal test facilities as opposed to developing new test facilities to meet the demands of a very fiscally constrained test and evaluation environment. The results of this study showed that Marshall Space Flight Center (MSFC) Test Stand (TS) 4550 could be made compliant, within reasonable cost and schedule impacts, if safety processes and operational limitations were put in place to meet the safety codes and concerns of the Fire Marshall. Trades were performed with key selection criteria to ensure that appropriate levels of occupant safety are incorporated into test facility design modifications. In preparation for the ground vibration tests that were to be performed on the Ares I launch vehicle, the Ares Flight and Integrated Test Office (FITO) organization evaluated the available test facility options, which included the existing mothballed structural dynamic TS4550 used by Apollo and Shuttle, alternative ground vibration test facilities at other locations, and construction of a new dynamic test stand. After an exhaustive assessment of the alternatives, the results favored modifying the TS4550 because it was the lowest cost option and presented the least schedule risk to the NASA Constellation Program for Ares Integrated Vehicle Ground Vibration Test (IVGVT). As the renovation design plans and drawings were being developed for TS4550, a safety concern was discovered the original design for the construction of the test stand, originally built for the Apollo Program and renovated for the Shuttle Program, was completed before NASA s adoption of the currently imposed safety and building codes per National Fire Protection Association Life Safety Code [NFPA 101] and International Building Codes. The initial FITO assessment of the design changes, required to make TS4550 compliant with current safety and building standards, identified a significant cost increase and schedule impact. An effort was launched to thoroughly evaluate the applicable life safety requirements, examine the context in which they were derived, and determine a means by which the TS4550 modifications could be made within budget and on schedule, while still providing the occupants with appropriate levels of safety.

Use the Bar Code System to Improve Accuracy of the Patient and Sample Identification.

PubMed

Chuang, Shu-Hsia; Yeh, Huy-Pzu; Chi, Kun-Hung; Ku, Hsueh-Chen

2018-01-01

In time and correct sample collection were highly related to patient's safety. The sample error rate was 11.1%, because misbranded patient information and wrong sample containers during January to April, 2016. We developed a barcode system of "Specimens Identify System" through process of reengineering of TRM, used bar code scanners, add sample container instructions, and mobile APP. Conclusion, the bar code systems improved the patient safety and created green environment.

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

Design of agricultural product quality safety retrospective supervision system of Jiangsu province

NASA Astrophysics Data System (ADS)

Wang, Kun

2017-08-01

In store and supermarkets to consumers can trace back agricultural products through the electronic province card to query their origin, planting, processing, packaging, testing and other important information and found that the problems. Quality and safety issues can identify the responsibility of the problem. This paper designs a retroactive supervision system for the quality and safety of agricultural products in Jiangsu Province. Based on the analysis of agricultural production and business process, the goal of Jiangsu agricultural product quality safety traceability system construction is established, and the specific functional requirements and non-functioning requirements of the retroactive system are analyzed, and the target is specified for the specific construction of the retroactive system. The design of the quality and safety traceability system in Jiangsu province contains the design of the overall design, the trace code design and the system function module.

2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The 2016 Annual Progress Report summarizes fiscal year 2016 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; market transformation; and Small Business Innovation Research projects.

2010 DOE Hydrogen Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

This report summarizes comments from the Peer Review Panel at the 2010 DOE Hydrogen Program Annual Merit Review, held on June 7-11, 2010, in Washington, DC. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; and systems analysis.

Learning health 'safety' within non-technical skills interprofessional simulation education: a qualitative study.

PubMed

Gordon, Morris; Fell, Christopher W R; Box, Helen; Farrell, Michael; Stewart, Alison

2017-01-01

Healthcare increasingly recognises and focusses on the phenomena of 'safe practice' and 'patient safety.' Success with non-technical skills (NTS) training in other industries has led to widespread transposition to healthcare education, with communication and teamwork skills central to NTS frameworks. This study set out to identify how the context of interprofessional simulation learning influences NTS acquisition and development of 'safety' amongst learners. Participants receiving a non-technical skills (NTS) safety focussed training package were invited to take part in a focus group interview which set out to explore communication, teamwork, and the phenomenon of safety in the context of the learning experiences they had within the training programme. The analysis was aligned with a constructivist paradigm and took an interactive methodological approach. The analysis proceeded through three stages, consisting of open, axial, and selective coding, with constant comparisons taking place throughout each phase. Each stage provided categories that could be used to explore the themes of the data. Additionally, to ensure thematic saturation, transcripts of observed simulated learning encounters were then analysed. Six themes were established at the axial coding level, i.e., analytical skills, personal behaviours, communication, teamwork, context, and pedagogy. Underlying these themes, two principal concepts emerged, namely: intergroup contact anxiety - as both a result of and determinant of communication - and teamwork, both of which must be considered in relation to context. These concepts have subsequently been used to propose a framework for NTS learning. This study highlights the role of intergroup contact anxiety and teamwork as factors in NTS behaviour and its dissipation through interprofessional simulation learning. Therefore, this should be a key consideration in NTS education. Future research is needed to consider the role of the affective non-technical attributes of intergroup contact anxiety and teamwork as focuses for education and determinants of safe behaviour. AUM: Anxiety/uncertainty management; NTS: Non-technical skills; TINSELS: Training in non-technical skills to enhance levels of medicines safety.

A comparison of IBC with 1997 UBC for modal response spectrum analysis in standard-occupancy buildings

NASA Astrophysics Data System (ADS)

Nahhas, Tariq M.

2011-03-01

This paper presents a comparison of the seismic forces generated from a Modal Response Spectrum Analysis (MRSA) by applying the provisions of two building codes, the 1997 Uniform Building Code (UBC) and the 2000-2009 International Building Code (IBC), to the most common ordinary residential buildings of standard occupancy. Considering IBC as the state of the art benchmark code, the primary concern is the safety of buildings designed using the UBC as compared to those designed using the IBC. A sample of four buildings with different layouts and heights was used for this comparison. Each of these buildings was assumed to be located at four different geographical sample locations arbitrarily selected to represent various earthquake zones on a seismic map of the USA, and was subjected to code-compliant response spectrum analyses for all sample locations and for five different soil types at each location. Response spectrum analysis was performed using the ETABS software package. For all the cases investigated, the UBC was found to be significantly more conservative than the IBC. The UBC design response spectra have higher spectral accelerations, and as a result, the response spectrum analysis provided a much higher base shear and moment in the structural members as compared to the IBC. The conclusion is that ordinary office and residential buildings designed using UBC 1997 are considered to be overdesigned, and therefore they are quite safe even according to the IBC provisions.

The application of the psychological contract to workplace safety.

PubMed

Walker, Arlene; Hutton, Dorothy M

2006-01-01

Psychological contracts of safety are conceptualized as the beliefs of individuals about reciprocal safety obligations inferred from implicit or explicit promises. Although the literature on psychological contracts is growing, the existence of psychological contracts in relation to safety has not been established. The research sought to identify psychological contracts in the conversations of employees about safety, by demonstrating reciprocity in relation to employer and employee safety obligations. The identified safety obligations were used to develop a measure of psychological contracts of safety. The participants were 131 employees attending safety training sessions in retail and manufacturing organizations. Non-participant observation was used to collect the data during safety training sessions. Content analysis was used to analyze the data. Categories for coding were established through identification of language markers that demonstrated contingencies or other implied obligations. Direct evidence of reciprocity between employer safety obligations and employee safety obligations was found in statements from the participants demonstrating psychological contracts. A comprehensive list of perceived employer and employee safety obligations was compiled and developed into a measure of psychological contracts of safety. A small sample of 33 safety personnel was used to validate the safety obligations. CONCLUSIONS AND IMPACT ON INDUSTRY: Implications of these findings for safety and psychological contract research are discussed.

42 CFR 137.368 - Is the Secretary responsible for oversight and compliance of health and safety codes during...

Code of Federal Regulations, 2010 CFR

2010-10-01

... compliance of health and safety codes during construction projects being performed by a Self-Governance Tribe... SERVICES TRIBAL SELF-GOVERNANCE Construction Roles of the Secretary in Establishing and Implementing Construction Project Agreements § 137.368 Is the Secretary responsible for oversight and compliance of health...

Probabilistic approach for decay heat uncertainty estimation using URANIE platform and MENDEL depletion code

NASA Astrophysics Data System (ADS)

Tsilanizara, A.; Gilardi, N.; Huynh, T. D.; Jouanne, C.; Lahaye, S.; Martinez, J. M.; Diop, C. M.

2014-06-01

The knowledge of the decay heat quantity and the associated uncertainties are important issues for the safety of nuclear facilities. Many codes are available to estimate the decay heat. ORIGEN, FISPACT, DARWIN/PEPIN2 are part of them. MENDEL is a new depletion code developed at CEA, with new software architecture, devoted to the calculation of physical quantities related to fuel cycle studies, in particular decay heat. The purpose of this paper is to present a probabilistic approach to assess decay heat uncertainty due to the decay data uncertainties from nuclear data evaluation like JEFF-3.1.1 or ENDF/B-VII.1. This probabilistic approach is based both on MENDEL code and URANIE software which is a CEA uncertainty analysis platform. As preliminary applications, single thermal fission of uranium 235, plutonium 239 and PWR UOx spent fuel cell are investigated.

Patient Health Goals Elicited During Home Care Admission: A Categorization.

PubMed

Sockolow, Paulina; Radhakrishnan, Kavita; Chou, Edgar Y; Wojciechowicz, Christine

2017-11-01

Home care agencies are initiating "patient health goal elicitation" activities as part of home care admission planning. We categorized elicited goals and identified "clinically informative" goals at a home care agency. We examined patient goals that admitting clinicians documented in the point-of-care electronic health record; conducted content analysis on patient goal data to develop a coding scheme; grouped goal themes into codes; assigned codes to each goal; and identified goals that were in the patient voice. Of the 1,763 patient records, 16% lacked a goal; only 15 goals were in a patient's voice. Nurse and physician experts identified 12 of the 20 codes as clinically important accounting for 82% of goal occurrences. The most frequent goal documented was safety/falls (23%). Training and consistent communication of the intent and operationalization of patient goal elicitation may address the absence of patient voice and the less than universal recording of home care patients' goals.

Department of Energy Construction Safety Reference Guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1993-09-01

DOE has adopted the Occupational Safety and Health Administration (OSHA) regulations Title 29 Code of Federal Regulations (CFR) 1926 ``Safety and Health Regulations for Construction,`` and related parts of 29 CFR 1910, ``Occupational Safety and Health Standards.`` This nonmandatory reference guide is based on these OSHA regulations and, where appropriate, incorporates additional standards, codes, directives, and work practices that are recognized and accepted by DOE and the construction industry. It covers excavation, scaffolding, electricity, fire, signs/barricades, cranes/hoists/conveyors, hand and power tools, concrete/masonry, stairways/ladders, welding/cutting, motor vehicles/mechanical equipment, demolition, materials, blasting, steel erection, etc.

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

SER assistant: An expert system for safety evaluation reports

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeChaine, M.D.; Levine, S.H.; Feltus, M.A.

1993-01-01

The SER Assistant is an expert system that assists engineers to write safety evaluation reports (SERs). Section 50.59 of the Code of Federal Regulations allows modifications to be made to nuclear power plants without prior US Nuclear Regulatory Commission approval if two conditions are satisfied. First, the change must not affect the technical specifications of the plant. Second, the modification must not affect a part of the plant described in the final safety analysis report, or if it does, it must not create an unreviewed safety question. The purpose of an SER is to ensure that these conditions are satisfiedmore » for the proposed modification. The SER Assistant aids this process by providing relevant, but directed, questions and information as well as giving engineers an organized environment to document their thought processes.« less

[Qualitative evaluation of employer requirements associated with occupational health and safety as good practice in small-scale enterprises].

PubMed

Kuroki, Naomi; Miyashita, Nana; Hino, Yoshiyuki; Kayashima, Kotaro; Fujino, Yoshihisa; Takada, Mikio; Nagata, Tomohisa; Yamataki, Hajime; Sakuragi, Sonoko; Kan, Hirohiko; Morita, Tetsuya; Ito, Akiyoshi; Mori, Koji

2009-09-01

The purpose of this study was to identify what motivates employers to promote good occupational health and safety practices in small-scale enterprises. Previous studies have shown that small-scale enterprises generally pay insufficient attention to issues of occupational health and safety. These findings were mainly derived from questionnaire based surveys. Nevertheless, some small-scale enterprises in which employers exercise good leadership do take a progressive approach to occupational health and safety. Although good practices can be identified in small-scale enterprises, it remains unclear what motivates employers in small-scale enterprises to actively implement occupational health and safety practices. We speculated that identifying employer motivations in promoting occupational health would help to spread good practices among small-scale enterprises. Using a qualitative approach based on the KJ methods, we interviewed ten employers who actively promote occupational health and safety in the workplace. The employers were asked to discuss their views of occupational health and safety in their own words. A semi-structured interview format was used, and transcripts were made of the interviews. Each transcript was independently coded by two or more researchers. These transcripts and codes were integrated and then the research group members discussed the heading titles and structural relationships between them according to the KJ method. Qualitative analysis revealed that all the employers expressed a strong interest in a "good company" and "good management". They emphasized four elements of "good management", namely "securing human resources", "trust of business partners", "social responsibility" and "employer's health condition itself", and considered that addressing occupational health and safety was essential to the achievement of these four elements. Consistent with previous findings, the results showed that implementation of occupational health and safety activities depended on "cost", "human resources", "time to perform", and "advisory organization". These results suggest that employer awareness of the relationship between good management and occupational health is essential to the implementation of occupational health and safety practices in small-scale enterprises.

Safety Education and Science.

ERIC Educational Resources Information Center

Ralph, Richard

1980-01-01

Safety education in the science classroom is discussed, including the beginning of safe management, attitudes toward safety education, laboratory assistants, chemical and health regulation, safety aids, and a case study of a high school science laboratory. Suggestions for safety codes for science teachers, student behavior, and laboratory…

Statechart Analysis with Symbolic PathFinder

NASA Technical Reports Server (NTRS)

Pasareanu, Corina S.

2012-01-01

We report here on our on-going work that addresses the automated analysis and test case generation for software systems modeled using multiple Statechart formalisms. The work is motivated by large programs such as NASA Exploration, that involve multiple systems that interact via safety-critical protocols and are designed with different Statechart variants. To verify these safety-critical systems, we have developed Polyglot, a framework for modeling and analysis of model-based software written using different Statechart formalisms. Polyglot uses a common intermediate representation with customizable Statechart semantics and leverages the analysis and test generation capabilities of the Symbolic PathFinder tool. Polyglot is used as follows: First, the structure of the Statechart model (expressed in Matlab Stateflow or Rational Rhapsody) is translated into a common intermediate representation (IR). The IR is then translated into Java code that represents the structure of the model. The semantics are provided as "pluggable" modules.

Implications of electronic health record downtime: an analysis of patient safety event reports.

PubMed

Larsen, Ethan; Fong, Allan; Wernz, Christian; Ratwani, Raj M

2018-02-01

We sought to understand the types of clinical processes, such as image and medication ordering, that are disrupted during electronic health record (EHR) downtime periods by analyzing the narratives of patient safety event report data. From a database of 80 381 event reports, 76 reports were identified as explicitly describing a safety event associated with an EHR downtime period. These reports were analyzed and categorized based on a developed code book to identify the clinical processes that were impacted by downtime. We also examined whether downtime procedures were in place and followed. The reports were coded into categories related to their reported clinical process: Laboratory, Medication, Imaging, Registration, Patient Handoff, Documentation, History Viewing, Delay of Procedure, and General. A majority of reports (48.7%, n = 37) were associated with lab orders and results, followed by medication ordering and administration (14.5%, n = 11). Incidents commonly involved patient identification and communication of clinical information. A majority of reports (46%, n = 35) indicated that downtime procedures either were not followed or were not in place. Only 27.6% of incidents (n = 21) indicated that downtime procedures were successfully executed. Patient safety report data offer a lens into EHR downtime-related safety hazards. Important areas of risk during EHR downtime periods were patient identification and communication of clinical information; these should be a focus of downtime procedure planning to reduce safety hazards. EHR downtime events pose patient safety hazards, and we highlight critical areas for downtime procedure improvement. © The Author 2017. Published by Oxford University Press on behalf of the American Medical Informatics Association. All rights reserved. For Permissions, please email: journals.permissions@oup.com

Assessment of tissue allograft safety monitoring with administrative healthcare databases: a pilot project using Medicare data.

PubMed

Dhakal, Sanjaya; Burwen, Dale R; Polakowski, Laura L; Zinderman, Craig E; Wise, Robert P

2014-03-01

Assess whether Medicare data are useful for monitoring tissue allograft safety and utilization. We used health care claims (billing) data from 2007 for 35 million fee-for-service Medicare beneficiaries, a predominantly elderly population. Using search terms for transplant-related procedures, we generated lists of ICD-9-CM and CPT(®) codes and assessed the frequency of selected allograft procedures. Step 1 used inpatient data and ICD-9-CM procedure codes. Step 2 added non-institutional provider (e.g., physician) claims, outpatient institutional claims, and CPT codes. We assembled preliminary lists of diagnosis codes for infections after selected allograft procedures. Many ICD-9-CM codes were ambiguous as to whether the procedure involved an allograft. Among 1.3 million persons with a procedure ascertained using the list of ICD-9-CM codes, only 1,886 claims clearly involved an allograft. CPT codes enabled better ascertainment of some allograft procedures (over 17,000 persons had corneal transplants and over 2,700 had allograft skin transplants). For spinal fusion procedures, CPT codes improved specificity for allografts; of nearly 100,000 patients with ICD-9-CM codes for spinal fusions, more than 34,000 had CPT codes indicating allograft use. Monitoring infrequent events (infections) after infrequent exposures (tissue allografts) requires large study populations. A strength of the large Medicare databases is the substantial number of certain allograft procedures. Limitations include lack of clinical detail and donor information. Medicare data can potentially augment passive reporting systems and may be useful for monitoring tissue allograft safety and utilization where codes clearly identify allograft use and coding algorithms can effectively screen for infections.

The Italian experience on T/H best estimate codes: Achievements and perspectives

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alemberti, A.; D`Auria, F.; Fiorino, E.

1997-07-01

Themalhydraulic system codes are complex tools developed to simulate the power plants behavior during off-normal conditions. Among the objectives of the code calculations the evaluation of safety margins, the operator training, the optimization of the plant design and of the emergency operating procedures, are mostly considered in the field of the nuclear safety. The first generation of codes was developed in the United States at the end of `60s. Since that time, different research groups all over the world started the development of their own codes. At the beginning of the `80s, the second generation codes were proposed; these differmore » from the first generation codes owing to the number of balance equations solved (six instead of three), the sophistication of the constitutive models and of the adopted numerics. The capabilities of available computers have been fully exploited during the years. The authors then summarize some of the major steps in the process of developing, modifying, and advancing the capabilities of the codes. They touch on the fact that Italian, and for that matter non-American, researchers have not been intimately involved in much of this work. They then describe the application of these codes in Italy, even though there are no operating or under construction nuclear power plants at this time. Much of this effort is directed at the general question of plant safety in the face of transient type events.« less

Proceedings of the Third International Workshop on Proof-Carrying Code and Software Certification

NASA Technical Reports Server (NTRS)

Ewen, Denney, W. (Editor); Jensen, Thomas (Editor)

2009-01-01

This NASA conference publication contains the proceedings of the Third International Workshop on Proof-Carrying Code and Software Certification, held as part of LICS in Los Angeles, CA, USA, on August 15, 2009. Software certification demonstrates the reliability, safety, or security of software systems in such a way that it can be checked by an independent authority with minimal trust in the techniques and tools used in the certification process itself. It can build on existing validation and verification (V&V) techniques but introduces the notion of explicit software certificates, Vvilich contain all the information necessary for an independent assessment of the demonstrated properties. One such example is proof-carrying code (PCC) which is an important and distinctive approach to enhancing trust in programs. It provides a practical framework for independent assurance of program behavior; especially where source code is not available, or the code author and user are unknown to each other. The workshop wiII address theoretical foundations of logic-based software certification as well as practical examples and work on alternative application domains. Here "certificate" is construed broadly, to include not just mathematical derivations and proofs but also safety and assurance cases, or any fonnal evidence that supports the semantic analysis of programs: that is, evidence about an intrinsic property of code and its behaviour that can be independently checked by any user, intermediary, or third party. These guarantees mean that software certificates raise trust in the code itself, distinct from and complementary to any existing trust in the creator of the code, the process used to produce it, or its distributor. In addition to the contributed talks, the workshop featured two invited talks, by Kelly Hayhurst and Andrew Appel. The PCC 2009 website can be found at http://ti.arc.nasa.gov /event/pcc 091.

A multicenter collaborative approach to reducing pediatric codes outside the ICU.

PubMed

Hayes, Leslie W; Dobyns, Emily L; DiGiovine, Bruno; Brown, Ann-Marie; Jacobson, Sharon; Randall, Kelly H; Wathen, Beth; Richard, Heather; Schwab, Carolyn; Duncan, Kathy D; Thrasher, Jodi; Logsdon, Tina R; Hall, Matthew; Markovitz, Barry

2012-03-01

The Child Health Corporation of America formed a multicenter collaborative to decrease the rate of pediatric codes outside the ICU by 50%, double the days between these events, and improve the patient safety culture scores by 5 percentage points. A multidisciplinary pediatric advisory panel developed a comprehensive change package of process improvement strategies and measures for tracking progress. Learning sessions, conference calls, and data submission facilitated collaborative group learning and implementation. Twenty Child Health Corporation of America hospitals participated in this 12-month improvement project. Each hospital identified at least 1 noncritical care target unit in which to implement selected elements of the change package. Strategies to improve prevention, detection, and correction of the deteriorating patient ranged from relatively simple, foundational changes to more complex, advanced changes. Each hospital selected a broad range of change package elements for implementation using rapid-cycle methodologies. The primary outcome measure was reduction in codes per 1000 patient days. Secondary outcomes were days between codes and change in patient safety culture scores. Code rate for the collaborative did not decrease significantly (3% decrease). Twelve hospitals reported additional data after the collaborative and saw significant improvement in code rates (24% decrease). Patient safety culture scores improved by 4.5% to 8.5%. A complex process, such as patient deterioration, requires sufficient time and effort to achieve improved outcomes and create a deeply embedded culture of patient safety. The collaborative model can accelerate improvements achieved by individual institutions.

46 CFR 12.15-3 - General requirements.

Code of Federal Regulations, 2011 CFR

2011-10-01

... years the minimum standards of competence for the following 4 areas of basic safety: (1) Personal... first aid as set out in table A-VI/1-3 of the STCW Code. (4) Personal safety and social responsibilities... of competence set out in STCW Regulation III/4 and Section A-III/4 of the STCW Code (incorporated by...

78 FR 28949 - Fund Availability Under VA's Homeless Providers Grant and Per Diem Program (Rehabilitation)

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-16

... Per Diem Program as a part of the effort to increase the useful life of the facilities of grantees... grantees are required to ensure that facilities rehabilitated under this NOFA meet the Life Safety Code of the National Fire and Protection Association. Please note, typically the Life Safety Code is more...

Surgical Site Infections Following Pediatric Ambulatory Surgery: An Epidemiologic Analysis.

PubMed

Rinke, Michael L; Jan, Dominique; Nassim, Janelle; Choi, Jaeun; Choi, Steven J

2016-08-01

OBJECTIVE To identify surgical site infection (SSI) rates following pediatric ambulatory surgery, SSI outcomes and risk factors, and sensitivity and specificity of SSI administrative billing codes. DESIGN Retrospective chart review of pediatric ambulatory surgeries with International Classification of Disease, Ninth Revision (ICD-9) codes for SSI, and a systematic random sampling of 5% of surgeries without SSI ICD-9 codes, all adjudicated for SSI on the basis of an ambulatory-adapted National Healthcare Safety Network definition. SETTING Urban pediatric tertiary care center April 1, 2009-March 31, 2014. METHODS SSI rates and sensitivity and specificity of ICD-9 codes were estimated using sampling design, and risk factors were analyzed in case-rest of cohort, and case-control, designs. RESULTS In 15,448 pediatric ambulatory surgeries, 34 patients had ICD-9 codes for SSI and 25 met the adapted National Healthcare Safety Network criteria. One additional SSI was identified with systematic random sampling. The SSI rate following pediatric ambulatory surgery was 2.9 per 1,000 surgeries (95% CI, 1.2-6.9). Otolaryngology surgeries demonstrated significantly lower SSI rates compared with endocrine (P=.001), integumentary (P=.001), male genital (P<.0001), and respiratory (P=.01) surgeries. Almost half of patients with an SSI were admitted, 88% received antibiotics, and 15% returned to the operating room. No risk factors were associated with SSI. The sensitivity of ICD-9 codes for SSI following ambulatory surgery was 55.31% (95% CI, 12.69%-91.33%) and specificity was 99.94% (99.89%-99.97%). CONCLUSIONS SSI following pediatric ambulatory surgery occurs at an appreciable rate and conveys morbidity on children. Infect Control Hosp Epidemiol 2016;37:931-938.

Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

NASA Astrophysics Data System (ADS)

Lobanov, P. D.; Usov, E. V.; Butov, A. A.; Pribaturin, N. A.; Mosunova, N. A.; Strizhov, V. F.; Chukhno, V. I.; Kutlimetov, A. E.

2017-10-01

Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, Friederike; Yoon, Su Jong

2015-01-01

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulationmore » requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference results. The status on the Monte Carlo modeling of the experimental VHTRC facility is also discussed. Reference results were obtained for the neutronics stand-alone cases (Ex. I-1 and Ex. I-2) using the (relatively new) Monte Carlo code Serpent, and comparisons were performed with the more established Monte Carlo codes MCNP and KENO-VI. For the thermal-fluids stand-alone cases (Ex. I-3 and I-4) the commercial CFD code CFX was utilized to obtain reference results that can be compared with lower fidelity tools.« less

78 FR 67048 - Prothioconazole; Pesticide Tolerances

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

... code 111). Animal production (NAICS code 112). Food manufacturing (NAICS code 311). Pesticide manufacturing (NAICS code 32532). B. How can I get electronic access to other related information? You may... Assessment and Determination of Safety Section 408(b)(2)(A)(i) of FFDCA allows EPA to establish a tolerance...

Improving radiopharmaceutical supply chain safety by implementing bar code technology.

PubMed

Matanza, David; Hallouard, François; Rioufol, Catherine; Fessi, Hatem; Fraysse, Marc

2014-11-01

The aim of this study was to describe and evaluate an approach for improving radiopharmaceutical supply chain safety by implementing bar code technology. We first evaluated the current situation of our radiopharmaceutical supply chain and, by means of the ALARM protocol, analysed two dispensing errors that occurred in our department. Thereafter, we implemented a bar code system to secure selected key stages of the radiopharmaceutical supply chain. Finally, we evaluated the cost of this implementation, from overtime, to overheads, to additional radiation exposure to workers. An analysis of the events that occurred revealed a lack of identification of prepared or dispensed drugs. Moreover, the evaluation of the current radiopharmaceutical supply chain showed that the dispensation and injection steps needed to be further secured. The bar code system was used to reinforce product identification at three selected key stages: at usable stock entry; at preparation-dispensation; and during administration, allowing to check conformity between the labelling of the delivered product (identity and activity) and the prescription. The extra time needed for all these steps had no impact on the number and successful conduct of examinations. The investment cost was reduced (2600 euros for new material and 30 euros a year for additional supplies) because of pre-existing computing equipment. With regard to the radiation exposure to workers there was an insignificant overexposure for hands with this new organization because of the labelling and scanning processes of radiolabelled preparation vials. Implementation of bar code technology is now an essential part of a global securing approach towards optimum patient management.

Improvements and applications of COBRA-TF for stand-alone and coupled LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, M.; Cuervo, D.; Ivanov, K.

2006-07-01

The advanced thermal-hydraulic subchannel code COBRA-TF has been recently improved and applied for stand-alone and coupled LWR core calculations at the Pennsylvania State Univ. in cooperation with AREVA NP GmbH (Germany)) and the Technical Univ. of Madrid. To enable COBRA-TF for academic and industrial applications including safety margins evaluations and LWR core design analyses, the code programming, numerics, and basic models were revised and substantially improved. The code has undergone through an extensive validation, verification, and qualification program. (authors)

Inventory of Safety-related Codes and Standards for Energy Storage Systems with some Experiences related to Approval and Acceptance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Conover, David R.

The purpose of this document is to identify laws, rules, model codes, codes, standards, regulations, specifications (CSR) related to safety that could apply to stationary energy storage systems (ESS) and experiences to date securing approval of ESS in relation to CSR. This information is intended to assist in securing approval of ESS under current CSR and to identification of new CRS or revisions to existing CRS and necessary supporting research and documentation that can foster the deployment of safe ESS.

Developing of a New Atmospheric Ionizing Radiation (AIR) Model

NASA Technical Reports Server (NTRS)

Clem, John M.; deAngelis, Giovanni; Goldhagen, Paul; Wilson, John W.

2003-01-01

As a result of the research leading to the 1998 AIR workshop and the subsequent analysis, the neutron issues posed by Foelsche et al. and further analyzed by Hajnal have been adequately resolved. We are now engaged in developing a new atmospheric ionizing radiation (AIR) model for use in epidemiological studies and air transportation safety assessment. A team was formed to examine a promising code using the basic FLUKA software but with modifications to allow multiple charged ion breakup effects. A limited dataset of the ER-2 measurements and other cosmic ray data will be used to evaluate the use of this code.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gharibyan, N.

In order to fully characterize the NIF neutron spectrum, SAND-II-SNL software was requested/received from the Radiation Safety Information Computational Center. The software is designed to determine the neutron energy spectrum through analysis of experimental activation data. However, given that the source code was developed in Sparcstation 10, it is not compatible with current version of FORTRAN. Accounts have been established through the Lawrence Livermore National Laboratory’s High Performance Computing in order to access different compiles for FORTRAN (e.g. pgf77, pgf90). Additionally, several of the subroutines included in the SAND-II-SNL package have required debugging efforts to allow for proper compiling ofmore » the code.« less

Benchmark Testing of a New 56Fe Evaluation for Criticality Safety Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leal, Luiz C; Ivanov, E.

2015-01-01

The SAMMY code was used to evaluate resonance parameters of the 56Fe cross section in the resolved resonance energy range of 0–2 MeV using transmission data, capture, elastic, inelastic, and double differential elastic cross sections. The resonance analysis was performed with the code SAMMY that fits R-matrix resonance parameters using the generalized least-squares technique (Bayes’ theory). The evaluation yielded a set of resonance parameters that reproduced the experimental data very well, along with a resonance parameter covariance matrix for data uncertainty calculations. Benchmark tests were conducted to assess the evaluation performance in benchmark calculations.

The science behind codes and standards for safe walkways: changes in level, stairways, stair handrails and slip resistance.

PubMed

Nemire, Kenneth; Johnson, Daniel A; Vidal, Keith

2016-01-01

Walkway codes and standards are often created through consensus by committees based on a number of factors, including historical precedence, common practice, cost, and empirical data. The authors maintain that in the formulation of codes and standards that impact pedestrian safety, the results of pertinent scientific research should be given significant weight. This article examines many elements of common walkway codes and standards related to changes in level, stairways, stair handrails, and slip resistance. It identifies which portions are based on or supported by empirical data; and which could benefit from additional scientific research. This article identifies areas in which additional research, codes, and standards may be beneficial to enhance pedestrian safety. Copyright © 2015 Elsevier Ltd and The Ergonomics Society. All rights reserved.

Poster - 28: Shielding of X-ray Rooms in Ontario in the Absence of Best Practice

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frimeth, Jeff; Richer, Jeff; Nesbitt, James

This poster will be strictly based on the Healing Arts Radiation Protection (HARP) Act, Regulation 543 under this Act (X-ray Safety Code), and personal communication the presenting author has had. In Ontario, the process of approval of an X-ray machine installation by the Director of the X-ray Inspection Service (XRIS) follows a certain protocol. Initially, the applicant submits a series of forms, including recommended shielding amounts, in order to satisfy the law. This documentation is then transferred to a third-party vendor (i.e. a professional engineer – P.Eng.) outsourced by the Ministry of Health and Long-term Care (MOHLTC). The P.Eng. thenmore » evaluates the submitted documentation for appropriate fulfillment of the HARP Act and Reg. 543 requirements. If the P.Eng.’s evaluation of the documentation is to their satisfaction, the XRIS is then notified. Finally, the Director will then issue a letter of approval to install the equipment at the facility. The methodology required to be used by the P.Eng. in order to determine the required amounts of protective barriers, and recommended to be used by the applicant, is contained within Safety Code 20A. However, Safety Code 35 has replaced the obsolete Safety Code 20A document and employs best practices in shielding design. This talk will focus further on specific intentions and limitations of Safety Code 20A. Furthermore, this talk will discuss the definition of the “practice of professional engineering” in Ontario. COMP members who are involved in shielding design are strongly encouraged to attend.« less

Hydrogen and Fuel Cell Technology | Transportation Research | NREL

Science.gov Websites

Outlines Safety Considerations for Hydrogen Technologies While safety requirements for industrial uses of vehicles have created the need for additional safety requirements. The new Hydrogen Technologies Safety hydrogen safety in context. For example, code officials reviewing permit applications for hydrogen projects

An Alignment Analysis of the U.S. Navy Supply Corps Officer’s Career Guidance With Naval Supply Systems Command’s Strategic Publications

DTIC Science & Technology

2014-06-03

relationship to business outcomes such as customer satisfaction , turnover, safety, and productivity” (Shuck, 2011, p. 312). Follow-on studies using...during the analysis efforts of this research. The 339 Level 1 codes span a wide range of ideas from strategy execution to customer satisfaction and...not reflect the radical shift in corporate culture needed by Naval Supply Systems Command (NAVSUP) to better serve U.S. Navy customers . This research

Scanning for safety: an integrated approach to improved bar-code medication administration.

PubMed

Early, Cynde; Riha, Chris; Martin, Jennifer; Lowdon, Karen W; Harvey, Ellen M

2011-03-01

This is a review of lessons learned in the postimplementation evaluation of a bar-code medication administration technology implemented at a major tertiary-care hospital in 2001. In 2006, with a bar-code medication administration scan compliance rate of 82%, a near-miss sentinel event prompted review of this technology as part of an institutional recommitment to a "culture of safety." Multifaceted problems with bar-code medication administration created an environment of circumventing safeguards as demonstrated by an increase in manual overrides to ensure timely medication administration. A multiprofessional team composed of nursing, pharmacy, human resources, quality, and technical services formalized. Each step in the bar-code medication administration process was reviewed. Technology, process, and educational solutions were identified and implemented systematically. Overall compliance with bar-code medication administration rose from 82% to 97%, which resulted in a calculated cost avoidance of more than $2.8 million during this time frame of the project.

Driver anger on the information superhighway: A content analysis of online complaints of offensive driver behaviour.

PubMed

Wickens, Christine M; Wiesenthal, David L; Hall, Ashley; Roseborough, James E W

2013-03-01

In recent years, several websites have been developed allowing drivers to post their complaints about other motorists online. These websites allow drivers to describe the nature of the offensive behaviour and to identify the offending motorist by vehicle type, colour, and license plate number. Some websites also ask drivers to list the location where the event took place and the exact date and time of the offence. The current study was a content analysis of complaints posted to RoadRagers.com between 1999 and 2007 (N=5624). The purpose of the study was to: (1) assess the research value of this novel data source; (2) demonstrate the value of content analysis to the study of driver behaviour; (3) further validate an existing coding scheme; (4) determine whether this new data source would replicate previous research findings regarding the most frequent types of driver complaints and temporal distribution of these reports; (5) provide recommendations for improved driver training and public safety initiatives based on these data. A coding scheme that was originally developed for an assessment of complaints submitted to the Ontario Provincial Police (OPP) (Wickens et al., 2005) was revised to accommodate the new dataset. The inter-rater reliability of the revised coding scheme as applied to the website complaints was very good (kappa=.85). The most frequently reported improper driver behaviours were cutting/weaving, speeding, perceived displays of hostility, and tailgating. Reports were most frequent on weekdays and during the morning and afternoon rush hour. The current study replicated several findings from the analysis of reports to the OPP, but possible differences in the sample and data collection method also produced some differences in findings. The value of content analysis to driver behaviour research and of driver complaint websites as a data source was demonstrated. Implications for driver safety initiatives and future research will be discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Safety with surgical lasers.

PubMed

McKenzie, A L

1984-01-01

As the sales of surgical lasers continue to grow, the problem of laser safety in hospitals becomes increasingly more urgent. This article considers both the principles and the practice of laser safety, and indicates how safety codes should be organized within a hospital. Eye safety is of paramount importance, and the effects of different wavelengths of laser radiation on the eye are described, both for intrabeam and extended-source exposure. An account is given of the concept of Maximum Permissible Exposure (MPE) and how it depends upon wavelength and exposure duration. The standard laser classification is developed in relation to MPE. The use of laser protective eyewear is discussed for the surgeon, other theatre staff and the patient. Finally, the role of the Laser Protection Supervisor and of the Laser Protection Adviser are explained in the context of establishing a local laser safety code.

The new interactive CESAR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, P.B.; Yatabe, M.

1987-01-01

In this report the Nuclear Criticality Safety Analytical Methods Resource Center describes a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) database at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR database was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE Criticality Safety Analytical Sequences and the codes comprising those sequences. The database includes data for both cross section preparation and criticality safety calculations. 3 refs., 1 tab.

New interactive CESAR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, P.B.; Yatabe, M.

1987-01-01

The Nuclear Criticality Safety Analytical Methods Resource Center announces the availability of a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) data base at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR data base was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE criticality safety analytical sequences and the codes comprising those sequences. The data base includes data for both cross-section preparation and criticality safety calculations.

Monte Carlo capabilities of the SCALE code system

DOE PAGES

Rearden, Bradley T.; Petrie, Jr., Lester M.; Peplow, Douglas E.; ...

2014-09-12

SCALE is a broadly used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play” framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport asmore » well as activation, depletion, and decay calculations. SCALE’s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE PAGES

Hu, Rui

2016-11-19

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

Post-licensure rapid immunization safety monitoring program (PRISM) data characterization.

PubMed

Baker, Meghan A; Nguyen, Michael; Cole, David V; Lee, Grace M; Lieu, Tracy A

2013-12-30

The Post-Licensure Rapid Immunization Safety Monitoring (PRISM) program is the immunization safety monitoring component of FDA's Mini-Sentinel project, a program to actively monitor the safety of medical products using electronic health information. FDA sought to assess the surveillance capabilities of this large claims-based distributed database for vaccine safety surveillance by characterizing the underlying data. We characterized data available on vaccine exposures in PRISM, estimated how much additional data was gained by matching with select state and local immunization registries, and compared vaccination coverage estimates based on PRISM data with other available data sources. We generated rates of computerized codes representing potential health outcomes relevant to vaccine safety monitoring. Standardized algorithms including ICD-9 codes, number of codes required, exclusion criteria and location of the encounter were used to obtain the background rates. The majority of the vaccines routinely administered to infants, children, adolescents and adults were well captured by claims data. Immunization registry data in up to seven states comprised between 5% and 9% of data for all vaccine categories with the exception of 10% for hepatitis B and 3% and 4% for rotavirus and zoster respectively. Vaccination coverage estimates based on PRISM's computerized data were similar to but lower than coverage estimates from the National Immunization Survey and Healthcare Effectiveness Data and Information Set. For the 25 health outcomes of interest studied, the rates of potential outcomes based on ICD-9 codes were generally higher than rates described in the literature, which are typically clinically confirmed cases. PRISM program's data on vaccine exposures and health outcomes appear complete enough to support robust safety monitoring. Copyright © 2013 Elsevier Ltd. All rights reserved.

Aluminum Data Measurements and Evaluation for Criticality Safety Applications

NASA Astrophysics Data System (ADS)

Leal, L. C.; Guber, K. H.; Spencer, R. R.; Derrien, H.; Wright, R. Q.

2002-12-01

The Defense Nuclear Facility Safety Board (DNFSB) Recommendation 93-2 motivated the US Department of Energy (DOE) to develop a comprehensive criticality safety program to maintain and to predict the criticality of systems throughout the DOE complex. To implement the response to the DNFSB Recommendation 93-2, a Nuclear Criticality Safety Program (NCSP) was created including the following tasks: Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data portion of the NCSP consists of a variety of differential measurements performed at the Oak Ridge Electron Linear Accelerator (ORELA) at the Oak Ridge National Laboratory (ORNL), data analysis and evaluation using the generalized least-squares fitting code SAMMY in the resolved, unresolved, and high energy ranges, and the development and benchmark testing of complete evaluations for a nuclide for inclusion into the Evaluated Nuclear Data File (ENDF/B). This paper outlines the work performed at ORNL to measure, evaluate, and test the nuclear data for aluminum for applications in criticality safety problems.

2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

2014-10-01

This report summarizes comments from the Peer Review Panel at the 2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 16-20, 2014, in Washington, DC. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

This report summarizes comments from the Peer Review Panel at the 2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 8-12, 2015, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

2012-09-01

This report summarizes comments from the Peer Review Panel at the 2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 14-18, 2012, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

none,

2011-09-01

This report summarizes comments from the Peer Review Panel at the 2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 9-13, 2011, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials,more » and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.« less

Do code of conduct audits improve chemical safety in garment factories? Lessons on corporate social responsibility in the supply chain from Fair Wear Foundation.

PubMed

Lindholm, Henrik; Egels-Zandén, Niklas; Rudén, Christina

2016-10-01

In managing chemical risks to the environment and human health in supply chains, voluntary corporate social responsibility (CSR) measures, such as auditing code of conduct compliance, play an important role. To examine how well suppliers' chemical health and safety performance complies with buyers' CSR policies and whether audited factories improve their performance. CSR audits (n = 288) of garment factories conducted by Fair Wear Foundation (FWF), an independent non-profit organization, were analyzed using descriptive statistics and statistical modeling. Forty-three per cent of factories did not comply with the FWF code of conduct, i.e. received remarks on chemical safety. Only among factories audited 10 or more times was there a significant increase in the number of factories receiving no remarks. Compliance with chemical safety requirements in garment supply chains is low and auditing is statistically correlated with improvements only at factories that have undergone numerous audits.

Do code of conduct audits improve chemical safety in garment factories? Lessons on corporate social responsibility in the supply chain from Fair Wear Foundation

PubMed Central

2016-01-01

Background In managing chemical risks to the environment and human health in supply chains, voluntary corporate social responsibility (CSR) measures, such as auditing code of conduct compliance, play an important role. Objectives To examine how well suppliers’ chemical health and safety performance complies with buyers’ CSR policies and whether audited factories improve their performance. Methods CSR audits (n = 288) of garment factories conducted by Fair Wear Foundation (FWF), an independent non-profit organization, were analyzed using descriptive statistics and statistical modeling. Results Forty-three per cent of factories did not comply with the FWF code of conduct, i.e. received remarks on chemical safety. Only among factories audited 10 or more times was there a significant increase in the number of factories receiving no remarks. Conclusions Compliance with chemical safety requirements in garment supply chains is low and auditing is statistically correlated with improvements only at factories that have undergone numerous audits. PMID:27611103

Advanced Vibration Analysis Tool Developed for Robust Engine Rotor Designs

NASA Technical Reports Server (NTRS)

Min, James B.

2005-01-01

The primary objective of this research program is to develop vibration analysis tools, design tools, and design strategies to significantly improve the safety and robustness of turbine engine rotors. Bladed disks in turbine engines always feature small, random blade-to-blade differences, or mistuning. Mistuning can lead to a dramatic increase in blade forced-response amplitudes and stresses. Ultimately, this results in high-cycle fatigue, which is a major safety and cost concern. In this research program, the necessary steps will be taken to transform a state-of-the-art vibration analysis tool, the Turbo- Reduce forced-response prediction code, into an effective design tool by enhancing and extending the underlying modeling and analysis methods. Furthermore, novel techniques will be developed to assess the safety of a given design. In particular, a procedure will be established for using natural-frequency curve veerings to identify ranges of operating conditions (rotational speeds and engine orders) in which there is a great risk that the rotor blades will suffer high stresses. This work also will aid statistical studies of the forced response by reducing the necessary number of simulations. Finally, new strategies for improving the design of rotors will be pursued.

Requirements for a multifunctional code architecture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiihonen, O.; Juslin, K.

1997-07-01

The present paper studies a set of requirements for a multifunctional simulation software architecture in the light of experiences gained in developing and using the APROS simulation environment. The huge steps taken in the development of computer hardware and software during the last ten years are changing the status of the traditional nuclear safety analysis software. The affordable computing power on the safety analysts table by far exceeds the possibilities offered to him/her ten years ago. At the same time the features of everyday office software tend to set standards to the way the input data and calculational results aremore » managed.« less

Development of photovoltaic array and module safety requirements

NASA Technical Reports Server (NTRS)

1982-01-01

Safety requirements for photovoltaic module and panel designs and configurations likely to be used in residential, intermediate, and large-scale applications were identified and developed. The National Electrical Code and Building Codes were reviewed with respect to present provisions which may be considered to affect the design of photovoltaic modules. Limited testing, primarily in the roof fire resistance field was conducted. Additional studies and further investigations led to the development of a proposed standard for safety for flat-plate photovoltaic modules and panels. Additional work covered the initial investigation of conceptual approaches and temporary deployment, for concept verification purposes, of a differential dc ground-fault detection circuit suitable as a part of a photovoltaic array safety system.

76 FR 22383 - National Fire Codes: Request for Proposals for Revision of Codes and Standards

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-21

... DEPARTMENT OF COMMERCE National Institute of Standards and Technology National Fire Codes: Request... publishing this notice on behalf of the National Fire Protection Association (NFPA) to announce the NFPA's proposal to revise some of its fire safety codes and standards and requests proposals from the public to...

44 CFR 206.400 - General.

Code of Federal Regulations, 2014 CFR

2014-10-01

... standards of safety, decency, and sanitation and in conformity with applicable codes, specifications and standards. (b) Applicable codes, specifications, and standards shall include any disaster resistant building code that meets the minimum requirements of the National Flood Insurance Program (NFIP) as well as...

44 CFR 206.400 - General.

Code of Federal Regulations, 2011 CFR

2011-10-01

... standards of safety, decency, and sanitation and in conformity with applicable codes, specifications and standards. (b) Applicable codes, specifications, and standards shall include any disaster resistant building code that meets the minimum requirements of the National Flood Insurance Program (NFIP) as well as...

44 CFR 206.400 - General.

Code of Federal Regulations, 2012 CFR

2012-10-01

... standards of safety, decency, and sanitation and in conformity with applicable codes, specifications and standards. (b) Applicable codes, specifications, and standards shall include any disaster resistant building code that meets the minimum requirements of the National Flood Insurance Program (NFIP) as well as...

44 CFR 206.400 - General.

Code of Federal Regulations, 2013 CFR

2013-10-01

... standards of safety, decency, and sanitation and in conformity with applicable codes, specifications and standards. (b) Applicable codes, specifications, and standards shall include any disaster resistant building code that meets the minimum requirements of the National Flood Insurance Program (NFIP) as well as...

44 CFR 206.400 - General.

Code of Federal Regulations, 2010 CFR

2010-10-01

... standards of safety, decency, and sanitation and in conformity with applicable codes, specifications and standards. (b) Applicable codes, specifications, and standards shall include any disaster resistant building code that meets the minimum requirements of the National Flood Insurance Program (NFIP) as well as...

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

NASA Astrophysics Data System (ADS)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Patient safety incidents in hospice care: observations from interdisciplinary case conferences.

PubMed

Oliver, Debra Parker; Demiris, George; Wittenberg-Lyles, Elaine; Gage, Ashley; Dewsnap-Dreisinger, Mariah L; Luetkemeyer, Jamie

2013-12-01

In the home hospice environment, issues arise every day presenting challenges to the safety, care, and quality of the dying experience. The literature pertaining to the safety challenges in this environment is limited. The study explored two research questions; 1) What types of patient safety incidents occur in the home hospice setting? 2) How many of these incidents are recognized by the hospice staff and/or the patient or caregiver as a patient safety incident? Video-recordings of hospice interdisciplinary team case conferences were reviewed and coded for patient safety incidents. Patient safety incidents were defined as any event or circumstance that could have resulted or did result in unnecessary harm to the patient or caregiver, or that could have resulted or did result in a negative impact on the quality of the dying experience for the patient. Codes for categories of patient safety incidents were based on the International Classification for Patient Safety. The setting for the study included two rural hospice programs in one Midwestern state in the United States. One hospice team had two separately functioning teams, the second hospice had three teams. 54 video-recordings were reviewed and coded. Patient safety incidents were identified that involved issues in clinical process, medications, falls, family or caregiving, procedural problems, documentation, psychosocial issues, administrative challenges and accidents. This study distinguishes categories of patient safety events that occur in home hospice care. Although the scope and definition of potential patient safety incidents in hospice is unique, the events observed in this study are similar to those observed with in other settings. This study identifies an operating definition and a potential classification for further research on patient safety incidents in hospice. Further research and consensus building of the definition of patient safety incidents and patient safety incidents in this setting is recommended.

Construction safety program for the National Ignition Facility, July 30, 1999 (NIF-0001374-OC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benjamin, D W

1999-07-30

These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and contractors/subcontractors. The General Rules-Code of Safe Practices shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S and H A-l that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Rules-Code of Safe Practices. (An interpreter must briefmore » those employees who do not speak or read English fluently.) In addition, all contractors and subcontractors shall adopt a written General Rules-Code of Safe Practices that relates to their operations. The General Rules-Code of Safe Practices must be posted at a conspicuous location at the job site office or be provided to each supervisory employee who shall have it readily available. Copies of the General Rules-Code of Safe Practices can also be included in employee safety pamphlets.« less

75 FR 59658 - Airworthiness Directives; SOCATA Model TBM 700 Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-28

... drive assemblies. The European Aviation Safety Agency (EASA), which is the Technical Agent for the... United States Code specifies the FAA's authority to issue rules on aviation safety. Subtitle I, section... Part 39 Air transportation, Aircraft, Aviation safety, Incorporation by reference, Safety. The Proposed...

Efficient Type Representation in TAL

NASA Technical Reports Server (NTRS)

Chen, Juan

2009-01-01

Certifying compilers generate proofs for low-level code that guarantee safety properties of the code. Type information is an essential part of safety proofs. But the size of type information remains a concern for certifying compilers in practice. This paper demonstrates type representation techniques in a large-scale compiler that achieves both concise type information and efficient type checking. In our 200,000-line certifying compiler, the size of type information is about 36% of the size of pure code and data for our benchmarks, the best result to the best of our knowledge. The type checking time is about 2% of the compilation time.

Workplace Safety and Health Topics: Safety & Prevention

MedlinePlus

... Health Records (EHRs) and Patient Work Information Engineering Controls Equipment Design in Mining Falls in the Workplace Green, Safe, and Healthy Jobs – Prevention through Design Hierarchy of Controls Industry and Occupation Coding and Support Logging Safety ...

An investigation of tritium transfer in reactor loops

NASA Astrophysics Data System (ADS)

Ilyasova, O. H.; Mosunova, N. A.

2017-09-01

The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

SPLASH program for three dimensional fluid dynamics with free surface boundaries

NASA Astrophysics Data System (ADS)

Yamaguchi, A.

1996-05-01

This paper describes a three dimensional computer program SPLASH that solves Navier-Stokes equations based on the Arbitrary Lagrangian Eulerian (ALE) finite element method. SPLASH has been developed for application to the fluid dynamics problems including the moving boundary of a liquid metal cooled Fast Breeder Reactor (FBR). To apply SPLASH code to the free surface behavior analysis, a capillary model using a cubic Spline function has been developed. Several sample problems, e.g., free surface oscillation, vortex shedding development, and capillary tube phenomena, are solved to verify the computer program. In the analyses, the numerical results are in good agreement with the theoretical value or experimental observance. Also SPLASH code has been applied to an analysis of a free surface sloshing experiment coupled with forced circulation flow in a rectangular tank. This is a simplified situation of the flow field in a reactor vessel of the FBR. The computational simulation well predicts the general behavior of the fluid flow inside and the free surface behavior. Analytical capability of the SPLASH code has been verified in this study and the application to more practical problems such as FBR design and safety analysis is under way.

Overview of the U.S. DOE Hydrogen Safety, Codes and Standards Program. Part 4: Hydrogen Sensors; Preprint

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buttner, William J.; Rivkin, Carl; Burgess, Robert

Hydrogen sensors are recognized as a critical element in the safety design for any hydrogen system. In this role, sensors can perform several important functions including indication of unintended hydrogen releases, activation of mitigation strategies to preclude the development of dangerous situations, activation of alarm systems and communication to first responders, and to initiate system shutdown. The functionality of hydrogen sensors in this capacity is decoupled from the system being monitored, thereby providing an independent safety component that is not affected by the system itself. The importance of hydrogen sensors has been recognized by DOE and by the Fuel Cellmore » Technologies Office's Safety and Codes Standards (SCS) program in particular, which has for several years supported hydrogen safety sensor research and development. The SCS hydrogen sensor programs are currently led by the National Renewable Energy Laboratory, Los Alamos National Laboratory, and Lawrence Livermore National Laboratory. The current SCS sensor program encompasses the full range of issues related to safety sensors, including development of advance sensor platforms with exemplary performance, development of sensor-related code and standards, outreach to stakeholders on the role sensors play in facilitating deployment, technology evaluation, and support on the proper selection and use of sensors.« less

An exploratory study of the relationship between socioeconomic status and motor vehicle safety features.

PubMed

Girasek, Deborah C; Taylor, Brett

2010-04-01

The purpose of this study was to assess the association between motor vehicle owners' socioeconomic status (SES) and the safety of their motor vehicles. Truncated vehicle identification numbers (VINs) were obtained from the Maryland Motor Vehicle Administration office. ZIP code-level income and educational data were assigned to each VIN. Software was used to identify safety-related vehicle characteristics including crash test rating, availability of electronic stability control and side impact air bags, age, and weight. Correlations and analyses of variance were performed to assess whether a ZIP code's median household income and educational level were associated with its proportion of registered vehicles with safety features. For 13 of the 16 correlations performed, SES was significantly associated with the availability of vehicle safety features in a direction that favored upper-income individuals. Vehicle weight was not associated with income or education. When ZIP codes were divided into median household income quintiles, their mean proportions of safety features also differed significantly, in the same direction, for availability of electronic stability control, side impact air bags, vehicle age, and crash test ratings. Safer motor vehicles appear to be distributed along socioeconomic lines, with lower income groups experiencing more risk. This previously unidentified mechanism of disparity merits further study and the attention of policy makers.

A Systematic Review of the Use of Social Media for Food Safety Risk Communication.

PubMed

Overbey, Katie N; Jaykus, Lee-Ann; Chapman, Benjamin J

2017-09-01

This article covers the current published literature related to the use of social media in food safety and infectious disease communication. The aim was to analyze literature recommendations and draw conclusions about how best to utilize social media for food safety risk communication going forward. A systematic literature review was conducted, and 24 articles were included for analysis. The inclusion criteria were (i) original peer-reviewed articles and (ii) primary focus on communication through social media about food safety and/or infectious diseases. Studies were coded for themes about social media applications, benefits, limitations, and best practices. Trust and personal beliefs were important drivers of social media use. The wide reach, immediacy, and information gathering capacities of social media were frequently cited benefits. Suggestions for social media best practices were inconsistent among studies, and study designs were highly variable. More evidence-based suggestions are needed to better establish guidelines for social media use in food safety and infectious disease risk communication. The information gleaned from this review can be used to create effective messages for shaping food safety behaviors.

An examination of some safety issues among commercial motorcyclists in Nigeria: a case study.

PubMed

Arosanyin, Godwin Tunde; Olowosulu, Adekunle Taiwo; Oyeyemi, Gafar Matanmi

2013-01-01

The reduction of road crashes and injuries among motorcyclists in Nigeria requires a system inquiry into some safety issues at pre-crash, crash and post-crash stages to guide action plans. This paper examines safety issues such as age restriction, motorcycle engine capacity, highway code awareness, licence holding, helmet usage, crash involvement, rescue and payment for treatment among commercial motorcyclists. The primary data derived from a structured questionnaire administered to 334 commercial motorcyclists in Samaru, Zaria were analysed using descriptive statistics and logistic regression technique. There was total compliance with age restriction and motorcycle engine capacity. About 41.8% of the operators were not aware of the existence of the highway code. The odds of licence holding increased with highway code awareness, education with above senior secondary as the reference category and earnings. The odds of crash involvement decreased with highway code awareness, earnings and mode of operation. About 84% of the motorcyclists did not use crash helmet, in spite of being aware of the benefit, and 65.4% of motorcycle crashes was found to be with other road users. The promotion of safety among motorcyclists therefore requires strict traffic law enforcement and modification of road design to segregate traffic and protect pedestrians.

Quick Response codes for surgical safety: a prospective pilot study.

PubMed

Dixon, Jennifer L; Smythe, William Roy; Momsen, Lara S; Jupiter, Daniel; Papaconstantinou, Harry T

2013-09-01

Surgical safety programs have been shown to reduce patient harm; however, there is variable compliance. The purpose of this study is to determine if innovative technology such as Quick Response (QR) codes can facilitate surgical safety initiatives. We prospectively evaluated the use of QR codes during the surgical time-out for 40 operations. Feasibility and accuracy were assessed. Perceptions of the current time-out process and the QR code application were evaluated through surveys using a 5-point Likert scale and binomial yes or no questions. At baseline (n = 53), survey results from the surgical team agreed or strongly agreed that the current time-out process was efficient (64%), easy to use (77%), and provided clear information (89%). However, 65% of surgeons felt that process improvements were needed. Thirty-seven of 40 (92.5%) QR codes scanned successfully, of which 100% were accurate. Three scan failures resulted from excessive curvature or wrinkling of the QR code label on the body. Follow-up survey results (n = 33) showed that the surgical team agreed or strongly agreed that the QR program was clearer (70%), easier to use (57%), and more accurate (84%). Seventy-four percent preferred the QR system to the current time-out process. QR codes accurately transmit patient information during the time-out procedure and are preferred to the current process by surgical team members. The novel application of this technology may improve compliance, accuracy, and outcomes. Copyright © 2013 Elsevier Inc. All rights reserved.

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 4 2013-10-01 2013-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 4 2014-10-01 2014-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 4 2012-10-01 2012-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 4 2011-10-01 2011-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Code of Federal Regulations, 2013 CFR

2013-01-01

... 7 Agriculture 12 2013-01-01 2013-01-01 false Voluntary National Model Building Codes E Exhibit E... National Model Building Codes The following documents address the health and safety aspects of buildings and related structures and are voluntary national model building codes as defined in § 1924.4(h)(2) of...

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Code of Federal Regulations, 2014 CFR

2014-01-01

... 7 Agriculture 12 2014-01-01 2013-01-01 true Voluntary National Model Building Codes E Exhibit E to... Model Building Codes The following documents address the health and safety aspects of buildings and related structures and are voluntary national model building codes as defined in § 1924.4(h)(2) of this...

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Code of Federal Regulations, 2012 CFR

2012-01-01

... 7 Agriculture 12 2012-01-01 2012-01-01 false Voluntary National Model Building Codes E Exhibit E... National Model Building Codes The following documents address the health and safety aspects of buildings and related structures and are voluntary national model building codes as defined in § 1924.4(h)(2) of...

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

Study on Design of High Efficiency and Light Weight Composite Propeller Blade for a Regional Turboprop Aircraft

NASA Astrophysics Data System (ADS)

Kong, Changduk; Lee, Kyungsun

2013-03-01

In this study, aerodynamic and structural design of the composite propeller blade for a regional turboprop aircraft is performed. The thin and wide chord propeller blade of high speed turboprop aircraft should have proper strength and stiffness to carry various kinds of loads such as high aerodynamic bending and twisting moments and centrifugal forces. Therefore the skin-spar-foam sandwich structure using high strength and stiffness carbon/epoxy composite materials is used to improve the lightness. A specific design procedure is proposed in this work as follows; firstly the aerodynamic configuration design, which is acceptable for the design requirements, is carried out using the in-house code developed by authors, secondly the structure design loads are determined through the aerodynamic load case analysis, thirdly the spar flange and the skin are preliminarily sized by consideration of major bending moments and shear forces using both the netting rule and the rule of mixture, and finally, the stress analysis is performed to confirm the structural safety and stability using finite element analysis commercial code, MSC. NASTRAN/PATRAN. Furthermore the additional analysis is performed to confirm the structural safety due to bird strike impact on the blade during flight operation using a commercial code, ANSYS. To realize the proposed propeller design, the prototype blades are manufactured by the following procedure; the carbon/epoxy composite fabric prepregs are laid up for skin and spar on a mold using the hand lay-up method and consolidated with a proper temperature and vacuum in the oven. To finalize the structural design, the full-scale static structural test is performed under the simulated aerodynamic loads using 3 point loading method. From the experimental results, it is found that the designed blade has a good structural integrity, and the measured results agree well with the analytical results as well.

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

DOE PAGES

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; ...

2016-09-23

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials,more » and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.« less

VICTORIA: A mechanistic model for radionuclide behavior in the reactor coolant system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schaperow, J.H.; Bixler, N.E.

1996-12-31

VICTORIA is the U.S. Nuclear Regulatory Commission`s (NRC`s) mechanistic, best-estimate code for analysis of fission product release from the core and subsequent transport in the reactor vessel and reactor coolant system. VICTORIA requires thermal-hydraulic data (i.e., temperatures, pressures, and velocities) as input. In the past, these data have been taken from the results of calculations from thermal-hydraulic codes such as SCDAP/RELAP5, MELCOR, and MAAP. Validation and assessment of VICTORIA 1.0 have been completed. An independent peer review of VICTORIA, directed by Brookhaven National Laboratory and supported by experts in the areas of fuel release, fission product chemistry, and aerosol physics,more » has been undertaken. This peer review, which will independently assess the code`s capabilities, is nearing completion with the peer review committee`s final report expected in Dec 1996. A limited amount of additional development is expected as a result of the peer review. Following this additional development, the NRC plans to release VICTORIA 1.1 and an updated and improved code manual. Future plans mainly involve use of the code for plant calculations to investigate specific safety issues as they arise. Also, the code will continue to be used in support of the Phebus experiments.« less

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hall, D.G.: Watkins, J.C.

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In additionmore » to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use.« less

A hydrogen-oxygen rocket engine coolant passage design program (RECOP) for fluid-cooled thrust chambers and nozzles

NASA Technical Reports Server (NTRS)

Tomsik, Thomas M.

1994-01-01

The design of coolant passages in regeneratively cooled thrust chambers is critical to the operation and safety of a rocket engine system. Designing a coolant passage is a complex thermal and hydraulic problem requiring an accurate understanding of the heat transfer between the combustion gas and the coolant. Every major rocket engine company has invested in the development of thrust chamber computer design and analysis tools; two examples are Rocketdyne's REGEN code and Aerojet's ELES program. In an effort to augment current design capabilities for government and industry, the NASA Lewis Research Center is developing a computer model to design coolant passages for advanced regeneratively cooled thrust chambers. The RECOP code incorporates state-of-the-art correlations, numerical techniques and design methods, certainly minimum requirements for generating optimum designs of future space chemical engines. A preliminary version of the RECOP model was recently completed and code validation work is in progress. This paper introduces major features of RECOP and compares the analysis to design points for the first test case engine; the Pratt & Whitney RL10A-3-3A thrust chamber.

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Numerical Propulsion System Simulation: A Common Tool for Aerospace Propulsion Being Developed

NASA Technical Reports Server (NTRS)

Follen, Gregory J.; Naiman, Cynthia G.

2001-01-01

The NASA Glenn Research Center is developing an advanced multidisciplinary analysis environment for aerospace propulsion systems called the Numerical Propulsion System Simulation (NPSS). This simulation is initially being used to support aeropropulsion in the analysis and design of aircraft engines. NPSS provides increased flexibility for the user, which reduces the total development time and cost. It is currently being extended to support the Aviation Safety Program and Advanced Space Transportation. NPSS focuses on the integration of multiple disciplines such as aerodynamics, structure, and heat transfer with numerical zooming on component codes. Zooming is the coupling of analyses at various levels of detail. NPSS development includes using the Common Object Request Broker Architecture (CORBA) in the NPSS Developer's Kit to facilitate collaborative engineering. The NPSS Developer's Kit will provide the tools to develop custom components and to use the CORBA capability for zooming to higher fidelity codes, coupling to multidiscipline codes, transmitting secure data, and distributing simulations across different platforms. These powerful capabilities will extend NPSS from a zero-dimensional simulation tool to a multifidelity, multidiscipline system-level simulation tool for the full life cycle of an engine.

Enhancing Public Helicopter Safety as a Component of Homeland Security

DTIC Science & Technology

2016-12-01

Risk Assessment Tool GPS Global Positioning System IFR instrument flight rules ILS instrument landing system IMC instrument meteorological...flight rules ( IFR ) flying and the lack of a pre-flight risk assessment. Pilot fatigue is a factor that appeared in two of the accident reports (New...three common factors that emerged from the qualitative analysis of coding: inadequate proficiency of IFR flying, lack of a pre- flight risk assessment

PharmARTS: terminology web services for drug safety data coding and retrieval.

PubMed

Alecu, Iulian; Bousquet, Cédric; Degoulet, Patrice; Jaulent, Marie-Christine

2007-01-01

MedDRA and WHO-ART are the terminologies used to encode drug safety reports. The standardisation achieved with these terminologies facilitates: 1) The sharing of safety databases; 2) Data mining for the continuous reassessment of benefit-risk ratio at national or international level or in the pharmaceutical industry. There is some debate about the capacity of these terminologies for retrieving case reports related to similar medical conditions. We have developed a resource that allows grouping similar medical conditions more effectively than WHO-ART and MedDRA. We describe here a software tool facilitating the use of this terminological resource thanks to an RDF framework with support for RDF Schema inferencing and querying. This tool eases coding and data retrieval in drug safety.

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

Types and patterns of safety concerns in home care: client and family caregiver perspectives

PubMed Central

Tong, Catherine E.; Sims-Gould, Joanie; Martin-Matthews, Anne

2016-01-01

Objective Drawing on interviews with home care clients and their family caregivers, we sought to understand how these individuals conceptualize safety in the provision and receipt of home care, how they promote safety in the home space and how their safety concerns differ from those of home support workers. Design In-depth, semi-structured interviews were conducted with clients and family caregivers. The analysis included topic and analytical coding of participants' verbatim accounts. Setting Interviews were completed in British Columbia, Canada. Participants Totally 82 clients and 55 caregivers participated. Results Clients and family caregivers identified three types of safety concerns: physical, spatial and interpersonal. These concerns are largely multi-dimensional and intersectional. We present a conceptual model of client and caregiver safety concerns. We also examine the factors that intensify and mitigate safety concerns in the home. Conclusions In spite of safety concerns, clients and family caregivers overwhelmingly prefer to receive care in the home setting. Spatial and physical concerns are the most salient. The financial burden of creating a safe care space should not be the client's alone to bear. The conceptualization and promotion of safety in home care must recognize the roles, responsibilities and perspectives of all of the actors involved, including workers, clients and their caregivers. PMID:26832159

Evidence Arguments for Using Formal Methods in Software Certification

NASA Technical Reports Server (NTRS)

Denney, Ewen W.; Pai, Ganesh

2013-01-01

We describe a generic approach for automatically integrating the output generated from a formal method/tool into a software safety assurance case, as an evidence argument, by (a) encoding the underlying reasoning as a safety case pattern, and (b) instantiating it using the data produced from the method/tool. We believe this approach not only improves the trustworthiness of the evidence generated from a formal method/tool, by explicitly presenting the reasoning and mechanisms underlying its genesis, but also provides a way to gauge the suitability of the evidence in the context of the wider assurance case. We illustrate our work by application to a real example-an unmanned aircraft system- where we invoke a formal code analysis tool from its autopilot software safety case, automatically transform the verification output into an evidence argument, and then integrate it into the former.

The Development of Dynamic Human Reliability Analysis Simulations for Inclusion in Risk Informed Safety Margin Characterization Frameworks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeffrey C. Joe; Diego Mandelli; Ronald L. Boring

2015-07-01

The United States Department of Energy is sponsoring the Light Water Reactor Sustainability program, which has the overall objective of supporting the near-term and the extended operation of commercial nuclear power plants. One key research and development (R&D) area in this program is the Risk-Informed Safety Margin Characterization pathway, which combines probabilistic risk simulation with thermohydraulic simulation codes to define and manage safety margins. The R&D efforts to date, however, have not included robust simulations of human operators, and how the reliability of human performance or lack thereof (i.e., human errors) can affect risk-margins and plant performance. This paper describesmore » current and planned research efforts to address the absence of robust human reliability simulations and thereby increase the fidelity of simulated accident scenarios.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunett, A. J.; Fanning, T. H.

The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such asmore » SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.« less

Current and anticipated uses of the CATHARE code at EDF and FRAMATOME

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gandrille, J.L.; Vacher, J.L.; Poizat, F.

1997-07-01

This paper presents current industrial applications of the CATHARE code in the fields of Safety Studies and Simulators where the code is intensively used by FRAMATOME, EDF and CEA, the development partners of CATHARE. Future needs in these fields are also recapitulated.

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Code of Federal Regulations, 2011 CFR

2011-01-01

... National Model Building Codes The following documents address the health and safety aspects of buildings... International, Inc., 4051 West Flossmoor Road, Country Club Hills, Illinois 60477. 2 Southern Building Code Congress International, Inc., 900 Montclair Road, Birmingham, Alabama 35213-1206. 3 International...

Medical students' perceptions of a novel institutional incident reporting system : A thematic analysis.

PubMed

Gordon, Morris; Parakh, Dillan

2017-10-01

Errors in healthcare are a major patient safety issue, with incident reporting a key solution. The incident reporting system has been integrated within a new medical curriculum, encouraging medical students to take part in this key safety process. The aim of this study was to describe the system and assess how students perceived the reporting system with regards to its role in enhancing safety. Employing a thematic analysis, this study used interviews with medical students at the end of the first year. Thematic indices were developed according to the information emerging from the data. Through open, axial and then selective stages of coding, an understanding of how the system was perceived was established. Analysis of the interview specified five core themes: (1) Aims of the incident reporting system; (2) internalized cognition of the system; (3) the impact of the reporting system; (4) threshold for reporting; (5) feedback on the systems operation. Selective analysis revealed three overriding findings: lack of error awareness and error wisdom as underpinned by key theoretical constructs, student support of the principle of safety, and perceptions of a blame culture. Students did not interpret reporting as a manner to support institutional learning and safety, rather many perceived it as a tool for a blame culture. The impact reporting had on students was unexpected and may give insight into how other undergraduates and early graduates interpret such a system. Future studies should aim to produce interventions that can support a reporting culture.

42 CFR 403.744 - Condition of participation: Life safety from fire.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 42 Public Health 2 2014-10-01 2014-10-01 false Condition of participation: Life safety from fire... safety from fire. (a) General. An RNHCI must meet the following conditions: (1) Except as otherwise... Safety Code of the National Fire Protection Association. The Director of the Office of the Federal...

Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

DOE PAGES

Maljovec, D.; Liu, S.; Wang, B.; ...

2015-07-14

Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

Development of a generalized perturbation theory method for sensitivity analysis using continuous-energy Monte Carlo methods

DOE PAGES

Perfetti, Christopher M.; Rearden, Bradley T.

2016-03-01

The sensitivity and uncertainty analysis tools of the ORNL SCALE nuclear modeling and simulation code system that have been developed over the last decade have proven indispensable for numerous application and design studies for nuclear criticality safety and reactor physics. SCALE contains tools for analyzing the uncertainty in the eigenvalue of critical systems, but cannot quantify uncertainty in important neutronic parameters such as multigroup cross sections, fuel fission rates, activation rates, and neutron fluence rates with realistic three-dimensional Monte Carlo simulations. A more complete understanding of the sources of uncertainty in these design-limiting parameters could lead to improvements in processmore » optimization, reactor safety, and help inform regulators when setting operational safety margins. A novel approach for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was recently explored as academic research and has been found to accurately and rapidly calculate sensitivity coefficients in criticality safety applications. The work presented here describes a new method, known as the GEAR-MC method, which extends the CLUTCH theory for calculating eigenvalue sensitivity coefficients to enable sensitivity coefficient calculations and uncertainty analysis for a generalized set of neutronic responses using high-fidelity continuous-energy Monte Carlo calculations. Here, several criticality safety systems were examined to demonstrate proof of principle for the GEAR-MC method, and GEAR-MC was seen to produce response sensitivity coefficients that agreed well with reference direct perturbation sensitivity coefficients.« less

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Camous, F.; Jacq, F.; Chatelard, P.

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

10 CFR 851.27 - Reference sources.

Code of Federal Regulations, 2013 CFR

2013-01-01

...) American Society of Mechanical Engineers (ASME), P.O. Box 2300 Fairfield, NJ 07007. Telephone: 800-843-2763... Electrical Code,” (2005). (5) NFPA 70E, “Standard for Electrical Safety in the Workplace,” (2004). (6... Engineers (ASME) Boilers and Pressure Vessel Code, sections I through XII including applicable Code Cases...

10 CFR 851.27 - Reference sources.

Code of Federal Regulations, 2014 CFR

2014-01-01

...) American Society of Mechanical Engineers (ASME), P.O. Box 2300 Fairfield, NJ 07007. Telephone: 800-843-2763... Electrical Code,” (2005). (5) NFPA 70E, “Standard for Electrical Safety in the Workplace,” (2004). (6... Engineers (ASME) Boilers and Pressure Vessel Code, sections I through XII including applicable Code Cases...

46 CFR 197.204 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

... Institute “Code for Pressure Piping, Power Piping.” ASME Code means the American Society of Mechanical Engineers “Boiler and Pressure Vessel Code.” ASME PVHO-1 means the ANSI/ASME standard “Safety Standard for Pressure Vessels for Human Occupancy.” ATA means a measure of pressure expressed in terms of atmosphere...

46 CFR 197.204 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-10-01

... Institute “Code for Pressure Piping, Power Piping.” ASME Code means the American Society of Mechanical Engineers “Boiler and Pressure Vessel Code.” ASME PVHO-1 means the ANSI/ASME standard “Safety Standard for Pressure Vessels for Human Occupancy.” ATA means a measure of pressure expressed in terms of atmosphere...

46 CFR 197.204 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-10-01

... Institute “Code for Pressure Piping, Power Piping.” ASME Code means the American Society of Mechanical Engineers “Boiler and Pressure Vessel Code.” ASME PVHO-1 means the ANSI/ASME standard “Safety Standard for Pressure Vessels for Human Occupancy.” ATA means a measure of pressure expressed in terms of atmosphere...

46 CFR 197.204 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-10-01

... Institute “Code for Pressure Piping, Power Piping.” ASME Code means the American Society of Mechanical Engineers “Boiler and Pressure Vessel Code.” ASME PVHO-1 means the ANSI/ASME standard “Safety Standard for Pressure Vessels for Human Occupancy.” ATA means a measure of pressure expressed in terms of atmosphere...

30 CFR 905.816 - Performance standards-Surface mining activities.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.817 - Performance standards-Underground mining activities.

Code of Federal Regulations, 2014 CFR

2014-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.817 - Performance standards-Underground mining activities.

Code of Federal Regulations, 2012 CFR

2012-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.816 - Performance standards-Surface mining activities.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.816 - Performance standards-Surface mining activities.

Code of Federal Regulations, 2012 CFR

2012-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.817 - Performance standards-Underground mining activities.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.816 - Performance standards-Surface mining activities.

Code of Federal Regulations, 2014 CFR

2014-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

30 CFR 905.817 - Peformance standards-Underground mining activities.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Quality Control Act, Cal. Pub. Res. Code section 13000 et seq.; the California Water Code section 1200 et seq.; the California Air Pollution Control Laws, Cal. Health & Safety Code section 39000 et seq.; the..., DEPARTMENT OF THE INTERIOR PROGRAMS FOR THE CONDUCT OF SURFACE MINING OPERATIONS WITHIN EACH STATE CALIFORNIA...

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 12 2010-01-01 2010-01-01 false Voluntary National Model Building Codes E Exhibit E... HOUSING SERVICE, RURAL BUSINESS-COOPERATIVE SERVICE, RURAL UTILITIES SERVICE, AND FARM SERVICE AGENCY... National Model Building Codes The following documents address the health and safety aspects of buildings...

38 CFR 39.63 - Architectural design standards.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Association Life Safety Code and Errata (NFPA 101), the 2003 edition of the NFPA 5000, Building Construction... section, all applicable local and State building codes and regulations must be observed. In areas not subject to local or State building codes, the recommendations contained in the 2003 edition of the NFPA...

About the necessity to manage events coded with MedDRA prior to statistical analysis: proposal of a strategy with application to a randomized clinical trial, ANRS 099 ALIZE.

PubMed

Journot, Valérie; Tabuteau, Sophie; Collin, Fidéline; Molina, Jean-Michel; Chene, Geneviève; Rancinan, Corinne

2008-03-01

Since 2003, the Medical Dictionary for Regulatory Activities (MedDRA) is the regulatory standard for safety report in clinical trials in the European Community. Yet, we found no published example of a practical experience for a scientifically oriented statistical analysis of events coded with MedDRA. We took advantage of a randomized trial in HIV-infected patients with MedDRA-coded events to explain the difficulties encountered during the events analysis and the strategy developed to report events consistently with trial-specific objectives. MedDRA has a rich hierarchical structure, which allows the grouping of coded terms into 5 levels, the highest being "System Organ Class" (SOC). Each coded term may be related to several SOCs, among which one primary SOC is defined. We developed a new general 5-step strategy to select a SOC as trial primary SOC, consistently with trial-specific objectives for this analysis. We applied it to the ANRS 099 ALIZE trial, where all events were coded with MedDRA version 3.0. We compared the MedDRA and the ALIZE primary SOCs. In the ANRS 099 ALIZE trial, 355 patients were recruited, and 3,722 events were reported and documented, among which 35% had multiple SOCs (2 to 4). We applied the proposed 5-step strategy. Altogether, 23% of MedDRA primary SOCs were modified, mainly from MedDRA primary SOCs "Investigations" (69%) and "Ear and labyrinth disorders" (6%), for the ALIZE primary SOCs "Hepatobiliary disorders" (35%), "Musculoskeletal and connective tissue disorders" (21%), and "Gastrointestinal disorders" (15%). MedDRA largely enhanced in size and complexity with versioning and the development of Standardized MedDRA Queries. Yet, statisticians should not systematically rely on primary SOCs proposed by MedDRA to report events. A simple general 5-step strategy to re-classify events consistently with the trial-specific objectives might be useful in HIV trials as well as in other fields.

Alternative Fuels Data Center

Science.gov Websites

Texas Commission on Environmental Quality (TCEQ). Exemptions apply for the following: vehicles with a idling. (Reference Texas Statutes, Health and Safety Code 382.0191; and Texas Administrative Code

The development of fire evaluation system for detention and correctional occupancies

NASA Astrophysics Data System (ADS)

Nelson, H. E.; Shibe, A. J.

1984-12-01

A fire safety evaluation system for detention and correctional occupancies was developed. It can be used for determining if a facility has fire safety equivalent to that obtained by meeting the requirement of a given code. The system was calibrated for use with proposed chapters for detention and correctional occupancies of the Life Safety Code (1985). There are separate sets of requirements for each of four use conditions: one for zoned egress, one for zoned impeded egress, one for impeded egress, and one for contained. Within each set, there are two levels of evaluation: one for partially sprinklered and nonsprinklered buildings, and one for totally sprinklered buildings.

Using narrative text and coded data to develop hazard scenarios for occupational injury interventions

PubMed Central

Lincoln, A; Sorock, G; Courtney, T; Wellman, H; Smith, G; Amoroso, P

2004-01-01

Objective: To determine whether narrative text in safety reports contains sufficient information regarding contributing factors and precipitating mechanisms to prioritize occupational back injury prevention strategies. Design, setting, subjects, and main outcome measures: Nine essential data elements were identified in narratives and coded sections of safety reports for each of 94 cases of back injuries to United States Army truck drivers reported to the United States Army Safety Center between 1987 and 1997. The essential elements of each case were used to reconstruct standardized event sequences. A taxonomy of the event sequences was then developed to identify common hazard scenarios and opportunities for primary interventions. Results: Coded data typically only identified five data elements (broad activity, task, event/exposure, nature of injury, and outcomes) while narratives provided additional elements (contributing factor, precipitating mechanism, primary source) essential for developing our taxonomy. Three hazard scenarios were associated with back injuries among Army truck drivers accounting for 83% of cases: struck by/against events during motor vehicle crashes; falls resulting from slips/trips or loss of balance; and overexertion from lifting activities. Conclusions: Coded data from safety investigations lacked sufficient information to thoroughly characterize the injury event. However, the combination of existing narrative text (similar to that collected by many injury surveillance systems) and coded data enabled us to develop a more complete taxonomy of injury event characteristics and identify common hazard scenarios. This study demonstrates that narrative text can provide the additional information on contributing factors and precipitating mechanisms needed to target prevention strategies. PMID:15314055

Assessment of Literature Related to Combustion Appliance Venting Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, V. H.; Less, B. D.; Singer, B. C.

In many residential building retrofit programs, air tightening to increase energy efficiency is often constrained by safety concerns with naturally vented combustion appliances. Tighter residential buildings more readily depressurize when exhaust equipment is operated, making combustion appliances more prone to backdraft or spill combustion exhaust into the living space. Several measures, such as installation guidelines, vent sizing codes, and combustion safety diagnostics, are in place with the intent to prevent backdrafting and combustion spillage, but the diagnostics conflict and the risk mitigation objective is inconsistent. This literature review summarizes the metrics and diagnostics used to assess combustion safety, documents theirmore » technical basis, and investigates their risk mitigations. It compiles information from the following: codes for combustion appliance venting and installation; standards and guidelines for combustion safety diagnostics; research evaluating combustion safety diagnostics; research investigating wind effects on building depressurization and venting; and software for simulating vent system performance.« less

Seatbelt and helmet depiction on the big screen blockbuster injury prevention messages?

PubMed

Cowan, John A; Dubosh, Nicole; Hadley, Craig

2009-03-01

Injuries from vehicle crashes are a major cause of death among American youth. Many of these injuries are worsened because of noncompliant safety practices. Messages delivered by mass media are omnipresent in young peoples' lives and influence their behavior patterns. In this investigation, we analyzed seat belt and helmet messages from a sample of top-grossing motion pictures with emphasis on scene context and character demographics. Content analysis of 50 top-grossing motion pictures for years 2000 to 2004, with coding for seat belt and helmet usage by trained media coders. In 48 of 50 movies (53% PG-13; 33% R; 10% PG; 4% G) with vehicle scenes, 518 scenes (82% car/truck; 7% taxi/limo; 7% motorcycle; 4% bicycle/skateboard) were coded. Overall, seat belt and helmet usage rates were 15.4% and 33.3%, respectively, with verbal indications for seat belt or helmet use found in 1.0% of scenes. Safety compliance rates varied by character race (18.3% white; 6.5% black; p = 0.036). No differences in compliance rates were noted for high-speed or unsafe vehicle operation. The injury rate for noncompliant characters involved in crashes was 10.7%. A regression model demonstrated black character race and escape scenes most predictive of noncompliant safety behavior. Safety compliance messages and images are starkly absent in top-grossing motion pictures resulting in, at worst, a deleterious effect on vulnerable populations and public health initiatives, and, at minimum, a lost opportunity to prevent injury and death. Healthcare providers should call on the motion picture industry to improve safety compliance messages and images in their products delivered for mass consumption.

Patient Reporting of Safety experiences in Organisational Care Transfers (PRoSOCT): a feasibility study of a patient reporting tool as a proactive approach to identifying latent conditions within healthcare systems

PubMed Central

Scott, Jason; Waring, Justin; Heavey, Emily; Dawson, Pamela

2014-01-01

Background It is increasingly recognised that patients can play a role in reporting safety incidents. Studies have tended to focus on patients within hospital settings, and on the reporting of patient safety incidents as defined within a medical model of safety. This study aims to determine the feasibility of collecting and using patient experiences of safety as a proactive approach to identifying latent conditions of safety as patients undergo organisational care transfers. Methods and analysis The study comprises three components: (1) patients’ experiences of safety relating to a care transfer, (2) patients’ receptiveness to reporting experiences of safety, (3) quality improvement using patient experiences of safety. (1) A safety survey and evaluation form will be distributed to patients discharged from 15 wards across four clinical areas (cardiac, care of older people, orthopaedics and stroke) over 1 year. Healthcare professionals involved in the care transfer will be provided with a regular summary of patient feedback. (2) Patients (n=36) who return an evaluation form will be sampled representatively based on the four clinical areas and interviewed about their experiences of healthcare and safety and completing the survey. (3) Healthcare professionals (n=75) will be invited to participate in semistructured interviews and focus groups to discuss their experiences with and perceptions of receiving and using patient feedback. Data analysis will explore the relationship between patient experiences of safety and other indicators and measures of quality and safety. Interview and focus group data will be thematically analysed and triangulated with all other data sources using a convergence coding matrix. Ethics and dissemination The study has been granted National Health Service (NHS) Research Ethics Committee approval. Patient experiences of safety will be disseminated to healthcare teams for the purpose of organisational development and quality improvement. Results will be disseminated to study participants as well as through peer-reviewed outputs. PMID:24833698

Analysis of the hydrological safety of dams combining two numerical tools: Iber and DualSPHysics

NASA Astrophysics Data System (ADS)

González-Cao, J.; García-Feal, O.; Domínguez, J. M.; Crespo, A. J. C.; Gómez-Gesteira, M.

2018-02-01

The upgrade of the hydrological safety of dams is a critical issue to avoid failures that can dramatically affect people and assets. This paper shows a numerical methodology to analyse the safety of the Belesar dam (NW, Spain) based on two different numerical codes. First, a mesh-based code named Iber, suited to deal with large 2-D domains, is used to simulate the impoundment. The initial conditions and the inlet provided to Iber correspond to the maximum water elevation and the maximum expected inflow to the impoundment defined in the technical specifications of the dam, which are associated to the more hazardous operation conditions of the dam. Iber provides information about the time needed for water to attain the crest of the dam when floodgates are closed. In addition, it also provides the velocity of discharge when gates are opened. Then, a mesh-free code named DualSPHysics, which is especially suited to deal with complex and violent 3-D flows, is used to reproduce the behaviour of one of the spillways of the dam starting from the results obtained with Iber, which are used as inlet conditions for DualSPHysics. The combined results of both model show that the left spillway can discharge the surplus of water associated to the maximum inflow to the reservoir if the gates of the spillways are opened before the overtopping of the dam was observed. In addition, water depth measured on the spillway is considerably lower than the lateral walls, preventing overtopping. Finally, velocities at different points of the spillway showed to be in good agreement with theoretical values.

30 CFR 57.19096 - Familiarity with signal code.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 57.19096 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Personnel... signals for cages, skips, and mantrips when persons or materials are being transported shall be familiar...

30 CFR 56.19096 - Familiarity with signal code.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 56.19096 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL MINES Personnel... signals for cages, skips, and mantrips when persons or materials are being transported shall be familiar...

29 CFR 1952.101 - Developmental schedule.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR... effective July 1, 1973. (b) Complete revision of all occupational safety and health codes as proposed within... budget. (e) Establishment of specific occupational safety and health goals by July 1, 1974. These goals...

29 CFR 1952.101 - Developmental schedule.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR... effective July 1, 1973. (b) Complete revision of all occupational safety and health codes as proposed within... budget. (e) Establishment of specific occupational safety and health goals by July 1, 1974. These goals...

Theory-Driven Models for Correcting Fight or Flight Imbalance in Gulf War Illness

DTIC Science & Technology

2011-09-01

testing on software • Performed static and dynamic analysis on safety code Research Interests To understand how the nervous system operates, how...dynamics of these systems to reset control of the HPA-immune axis to normal. We have completed the negotiation of sub-awards to the CFIDS Association...We propose that severe physical or psychological insult to the endocrine and immune systems can displace these from a normal regulatory equilibrium

Current implementation and future plans on new code architecture, programming language and user interface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brun, B.

1997-07-01

Computer technology has improved tremendously during the last years with larger media capacity, memory and more computational power. Visual computing with high-performance graphic interface and desktop computational power have changed the way engineers accomplish everyday tasks, development and safety studies analysis. The emergence of parallel computing will permit simulation over a larger domain. In addition, new development methods, languages and tools have appeared in the last several years.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farren Hunt

Idaho National Laboratory (INL) performed an Annual Effectiveness Review of the Integrated Safety Management System (ISMS), per 48 Code of Federal Regulations (CFR) 970.5223 1, “Integration of Environment, Safety and Health into Work Planning and Execution.” The annual review assessed Integrated Safety Management (ISM) effectiveness, provided feedback to maintain system integrity, and identified target areas for focused improvements and assessments for fiscal year (FY) 2013. Results of the FY 2012 annual effectiveness review demonstrated that the INL’s ISMS program was significantly strengthened. Actions implemented by the INL demonstrate that the overall Integrated Safety Management System is sound and ensures safemore » and successful performance of work while protecting workers, the public, and environment. This report also provides several opportunities for improvement that will help further strengthen the ISM Program and the pursuit of safety excellence. Demonstrated leadership and commitment, continued surveillance, and dedicated resources have been instrumental in maturing a sound ISMS program. Based upon interviews with personnel, reviews of assurance activities, and analysis of ISMS process implementation, this effectiveness review concludes that ISM is institutionalized and is “Effective”.« less

Safety outcomes in the United States: trends and challenges in measurement.

PubMed

Greenberg, Michael D; Haviland, Amelia M; Yu, Hao; Farley, Donna O

2009-04-01

To prepare Agency for Healthcare Research and Quality (AHRQ) for monitoring the impact of its own patient safety initiative, by exploring available outcomes data, assessing usability of measures, and estimating national trends in patient outcomes. Annual summary data on incidence of Joint Commission Sentinel Events, MEDMARX medication error events, and MDS measures of falls and pressure ulcers in nursing home residents. HCUP National Inpatient Sample (NIS) administrative claims data. Description and assessment of published summary data on selected safety measures. Analysis of selected Patient Safety Indicators (PSIs) and Utah–Missouri adverse event measures using HCUP NIS claims data (1994–2003). Interpretation of safety outcome trends requires close attention to the characteristics of underlying data sources and measures. Encounter-based measures have been affected by changes in definitions and ICD-9 coding, as well as by changes in the structure of administrative datasets like HCUP NIS. Historical trends are mixed for the safety outcome measures reviewed, with some measures showing improvement, others deterioration, and still others remaining fairly stable. Constructing national trends of safety outcomes is difficult because of limitations in available data sources and measures. Tracking growth in the adoption of safe practices could offer an important strategy for complementing existing safety measurement capabilities.

36 CFR 910.37 - Fire and life safety.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 36 Parks, Forests, and Public Property 3 2011-07-01 2011-07-01 false Fire and life safety. 910.37... DEVELOPMENT AREA Standards Uniformly Applicable to the Development Area § 910.37 Fire and life safety. As a... recommended that all new development be guided by standards of the NFPA Codes for fire and life safety and...

36 CFR 910.37 - Fire and life safety.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 36 Parks, Forests, and Public Property 3 2010-07-01 2010-07-01 false Fire and life safety. 910.37... DEVELOPMENT AREA Standards Uniformly Applicable to the Development Area § 910.37 Fire and life safety. As a... recommended that all new development be guided by standards of the NFPA Codes for fire and life safety and...

Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical modelmore » in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Validation of a common data model for active safety surveillance research

PubMed Central

Ryan, Patrick B; Reich, Christian G; Hartzema, Abraham G; Stang, Paul E

2011-01-01

Objective Systematic analysis of observational medical databases for active safety surveillance is hindered by the variation in data models and coding systems. Data analysts often find robust clinical data models difficult to understand and ill suited to support their analytic approaches. Further, some models do not facilitate the computations required for systematic analysis across many interventions and outcomes for large datasets. Translating the data from these idiosyncratic data models to a common data model (CDM) could facilitate both the analysts' understanding and the suitability for large-scale systematic analysis. In addition to facilitating analysis, a suitable CDM has to faithfully represent the source observational database. Before beginning to use the Observational Medical Outcomes Partnership (OMOP) CDM and a related dictionary of standardized terminologies for a study of large-scale systematic active safety surveillance, the authors validated the model's suitability for this use by example. Validation by example To validate the OMOP CDM, the model was instantiated into a relational database, data from 10 different observational healthcare databases were loaded into separate instances, a comprehensive array of analytic methods that operate on the data model was created, and these methods were executed against the databases to measure performance. Conclusion There was acceptable representation of the data from 10 observational databases in the OMOP CDM using the standardized terminologies selected, and a range of analytic methods was developed and executed with sufficient performance to be useful for active safety surveillance. PMID:22037893

Uncertainty Quantification Techniques of SCALE/TSUNAMI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T; Mueller, Don

2011-01-01

The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system developed at Oak Ridge National Laboratory (ORNL) includes Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI). The TSUNAMI code suite can quantify the predicted change in system responses, such as k{sub eff}, reactivity differences, or ratios of fluxes or reaction rates, due to changes in the energy-dependent, nuclide-reaction-specific cross-section data. Where uncertainties in the neutron cross-section data are available, the sensitivity of the system to the cross-section data can be applied to propagate the uncertainties in the cross-section data to an uncertainty in the system response. Uncertainty quantification ismore » useful for identifying potential sources of computational biases and highlighting parameters important to code validation. Traditional validation techniques often examine one or more average physical parameters to characterize a system and identify applicable benchmark experiments. However, with TSUNAMI correlation coefficients are developed by propagating the uncertainties in neutron cross-section data to uncertainties in the computed responses for experiments and safety applications through sensitivity coefficients. The bias in the experiments, as a function of their correlation coefficient with the intended application, is extrapolated to predict the bias and bias uncertainty in the application through trending analysis or generalized linear least squares techniques, often referred to as 'data adjustment.' Even with advanced tools to identify benchmark experiments, analysts occasionally find that the application models include some feature or material for which adequately similar benchmark experiments do not exist to support validation. For example, a criticality safety analyst may want to take credit for the presence of fission products in spent nuclear fuel. In such cases, analysts sometimes rely on 'expert judgment' to select an additional administrative margin to account for gap in the validation data or to conclude that the impact on the calculated bias and bias uncertainty is negligible. As a result of advances in computer programs and the evolution of cross-section covariance data, analysts can use the sensitivity and uncertainty analysis tools in the TSUNAMI codes to estimate the potential impact on the application-specific bias and bias uncertainty resulting from nuclides not represented in available benchmark experiments. This paper presents the application of methods described in a companion paper.« less

Comparison of the LLNL ALE3D and AKTS Thermal Safety Computer Codes for Calculating Times to Explosion in ODTX and STEX Thermal Cookoff Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wemhoff, A P; Burnham, A K

2006-04-05

Cross-comparison of the results of two computer codes for the same problem provides a mutual validation of their computational methods. This cross-validation exercise was performed for LLNL's ALE3D code and AKTS's Thermal Safety code, using the thermal ignition of HMX in two standard LLNL cookoff experiments: the One-Dimensional Time to Explosion (ODTX) test and the Scaled Thermal Explosion (STEX) test. The chemical kinetics model used in both codes was the extended Prout-Tompkins model, a relatively new addition to ALE3D. This model was applied using ALE3D's new pseudospecies feature. In addition, an advanced isoconversional kinetic approach was used in the AKTSmore » code. The mathematical constants in the Prout-Tompkins code were calibrated using DSC data from hermetically sealed vessels and the LLNL optimization code Kinetics05. The isoconversional kinetic parameters were optimized using the AKTS Thermokinetics code. We found that the Prout-Tompkins model calculations agree fairly well between the two codes, and the isoconversional kinetic model gives very similar results as the Prout-Tompkins model. We also found that an autocatalytic approach in the beta-delta phase transition model does affect the times to explosion for some conditions, especially STEX-like simulations at ramp rates above 100 C/hr, and further exploration of that effect is warranted.« less

Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A.

1995-12-31

In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEUmore » codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.« less

Revenue sources for financing transportation safety activities in Virginia : phase two, state sources.

DOT National Transportation Integrated Search

1980-01-01

Senate Bill 85, an action of the 1978 General Assembly, amended the Code of Virginia to provide, in part, that the Division of Highway Safety be succeeded by the newly created Department of Transportation Safety effective July 1, 1978. In its Declara...

29 CFR 1910.36 - Design and construction requirements for exit routes.

Code of Federal Regulations, 2012 CFR

2012-07-01

....36 Section 1910.36 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR OCCUPATIONAL SAFETY AND HEALTH STANDARDS Exit Routes and Emergency Planning... of exit routes necessary for your workplace, consult NFPA 101-2000, Life Safety Code. (c) Exit...

Dynamic response analysis of a 24-story damped steel structure

NASA Astrophysics Data System (ADS)

Feng, Demin; Miyama, Takafumi

2017-10-01

In Japanese and Chinese building codes, a two-stage design philosophy, damage limitation (small earthquake, Level 1) and life safety (extreme large earthquake, Level 2), is adopted. It is very interesting to compare the design method of a damped structure based on the two building codes. In the Chinese code, in order to be consistent with the conventional seismic design method, the damped structure is also designed at the small earthquake level. The effect of damper systems is considered by the additional damping ratio concept. The design force will be obtained from the damped design spectrum considering the reduction due to the additional damping ratio. The additional damping ratio by the damper system is usually calculated by a time history analysis method at the small earthquake level. The velocity dependent type dampers such as viscous dampers can function well even in the small earthquake level. But, if steel damper is used, which usually remains elastic in the small earthquake, there will be no additional damping ratio achieved. On the other hand, a time history analysis is used in Japan both for small earthquake and extreme large earthquake level. The characteristics of damper system and ductility of the structure can be modelled well. An existing 24-story steel frame is modified to demonstrate the design process of the damped structure based on the two building codes. Viscous wall type damper and low yield steel panel dampers are studied as the damper system.

Three-dimensional analysis of tokamaks and stellarators

PubMed Central

Garabedian, Paul R.

2008-01-01

The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

Design-Load Basis for LANL Structures, Systems, and Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

I. Cuesta

2004-09-01

This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loadsmore » not related to natural phenomena hazards, and (3) the design loads on structures during construction.« less

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

Nimbus/TOMS Science Data Operations Support

NASA Technical Reports Server (NTRS)

Childs, Jeff

1998-01-01

1. Participate in and provide analysis of laboratory and in-flight calibration of UV sensors used for space observations of backscattered UV radiation. 2. Provide support to the TOMS Science Operations Center, including generating instrument command lists and analysis of TOMS health and safety data. 3. Develop and maintain software and algorithms designed to capture and process raw spacecraft and instrument data, convert the instrument output into measured radiance and irradiances, and produce scientifically valid products. 4. Process the TOMS data into Level 1, Level 2, and Level 3 data products. 5. Provide analysis of the science data products in support of NASA GSFC Code 916's research.

Nimbus/TOMS Science Data Operations Support

NASA Technical Reports Server (NTRS)

1998-01-01

Projected goals include the following: (1) Participate in and provide analysis of laboratory and in-flight calibration of LTV sensors used for space observations of backscattered LTV radiation; (2) Provide support to the TOMS Science Operations Center, including generating instrument command lists and analysis of TOMS health and safety data; (3) Develop and maintain software and algorithms designed to capture and process raw spacecraft and instrument data, convert the instrument output into measured radiance and irradiances, and produce scientifically valid products; (4) Process the TOMS data into Level 1, Level 2, and Level 3 data products; (5) Provide analysis of the science data products in support of NASA GSFC Code 916's research.

Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis (LDRD 17-0958) FY17 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, Timothy A.; Denman, Matthew R.; Williams, R. A.

Instrumentation and control of nuclear power is transforming from analog to modern digital assets. These control systems perform key safety and security functions. This transformation is occurring in new plant designs as well as in the existing fleet of plants as the operation of those plants is extended to 60 years. This transformation introduces new and unknown issues involving both digital asset induced safety issues and security issues. Traditional nuclear power risk assessment tools and cyber security assessment methods have not been modified or developed to address the unique nature of cyber failure modes and of cyber security threat vulnerabilities.more » iii This Lab-Directed Research and Development project has developed a dynamic cyber-risk in- formed tool to facilitate the analysis of unique cyber failure modes and the time sequencing of cyber faults, both malicious and non-malicious, and impose those cyber exploits and cyber faults onto a nuclear power plant accident sequence simulator code to assess how cyber exploits and cyber faults could interact with a plants digital instrumentation and control (DI&C) system and defeat or circumvent a plants cyber security controls. This was achieved by coupling an existing Sandia National Laboratories nuclear accident dynamic simulator code with a cyber emulytics code to demonstrate real-time simulation of cyber exploits and their impact on automatic DI&C responses. Studying such potential time-sequenced cyber-attacks and their risks (i.e., the associated impact and the associated degree of difficulty to achieve the attack vector) on accident management establishes a technical risk informed framework for developing effective cyber security controls for nuclear power.« less

A Generic Software Safety Document Generator

NASA Technical Reports Server (NTRS)

Denney, Ewen; Venkatesan, Ram Prasad

2004-01-01

Formal certification is based on the idea that a mathematical proof of some property of a piece of software can be regarded as a certificate of correctness which, in principle, can be subjected to external scrutiny. In practice, however, proofs themselves are unlikely to be of much interest to engineers. Nevertheless, it is possible to use the information obtained from a mathematical analysis of software to produce a detailed textual justification of correctness. In this paper, we describe an approach to generating textual explanations from automatically generated proofs of program safety, where the proofs are of compliance with an explicit safety policy that can be varied. Key to this is tracing proof obligations back to the program, and we describe a tool which implements this to certify code auto-generated by AutoBayes and AutoFilter, program synthesis systems under development at the NASA Ames Research Center. Our approach is a step towards combining formal certification with traditional certification methods.

Automated Analysis of Stateflow Models

NASA Technical Reports Server (NTRS)

Bourbouh, Hamza; Garoche, Pierre-Loic; Garion, Christophe; Gurfinkel, Arie; Kahsaia, Temesghen; Thirioux, Xavier

2017-01-01

Stateflow is a widely used modeling framework for embedded and cyber physical systems where control software interacts with physical processes. In this work, we present a framework a fully automated safety verification technique for Stateflow models. Our approach is two-folded: (i) we faithfully compile Stateflow models into hierarchical state machines, and (ii) we use automated logic-based verification engine to decide the validity of safety properties. The starting point of our approach is a denotational semantics of State flow. We propose a compilation process using continuation-passing style (CPS) denotational semantics. Our compilation technique preserves the structural and modal behavior of the system. The overall approach is implemented as an open source toolbox that can be integrated into the existing Mathworks Simulink Stateflow modeling framework. We present preliminary experimental evaluations that illustrate the effectiveness of our approach in code generation and safety verification of industrial scale Stateflow models.

Exploring approaches to patient safety: the case of spinal manipulation therapy.

PubMed

Rozmovits, Linda; Mior, Silvano; Boon, Heather

2016-06-02

The purpose of this study was to gain insight into the current safety culture around the use of spinal manipulation therapy (SMT) by regulated health professionals in Canada and to explore perceptions of readiness for implementing formal mechanisms for tracking associated adverse events. Fifty-six semi-structured telephone interviews were conducted with professional leaders and frontline practitioners in chiropractic, physiotherapy, naturopathy and medicine, all professions regulated to perform SMT in the provinces of Alberta and Ontario Canada. Interviews were digitally audio-recorded for verbatim transcription. Transcripts were entered into HyperResearch software for qualitative data analysis and were coded for both anticipated and emergent themes using the constant comparative method. A thematic, descriptive analysis was produced. The safety culture around SMT is characterized by substantial disagreement about its actual rather than putative risks. Competing intra- and inter-professional narratives further cloud the safety picture. Participants felt that safety talk is sometimes conflated with competition for business in the context of fee-for-service healthcare delivery by several professions with overlapping scopes of practice. Both professional leaders and frontline practitioners perceived multiple barriers to the implementation of an incident reporting system for SMT. The established 'measure and manage' approach to patient safety is difficult to apply to care which is geographically dispersed and delivered by practitioners in multiple professions with overlapping scopes of practice, primarily in a fee-for-service model. Collaboration across professions on models that allow practitioners to share information anonymously and help practitioners learn from the reported incidents is needed.

The Crash Outcome Data Evaluation System (CODES)

DOT National Transportation Integrated Search

1996-01-01

The CODES Technical Report presents state-specific results from the Crash : Outcome Data Evaluation System project. These results confirm previous NHTSA : studies and show that safety belts and motorcycle helmets are effective in : reducing fatalitie...

Project : semi-autonomous parking for enhanced safety and efficiency.

DOT National Transportation Integrated Search

2016-04-01

Index coding, a coding formulation traditionally analyzed in the theoretical computer science and : information theory communities, has received considerable attention in recent years due to its value in : wireless communications and networking probl...

Alternative Fuels Data Center

Science.gov Websites

Propane and Natural Gas Safety The Railroad Commission of Texas regulates the safety of the natural gas and propane industries. (Reference Texas Statutes, Natural Resources Code 113.011 and 116.011

The role of the uncertainty in code development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barre, F.

1997-07-01

From a general point of view, all the results of a calculation should be given with their uncertainty. It is of most importance in nuclear safety where sizing of the safety systems, therefore protection of the population and the environment essentially depends on the calculation results. Until these last years, the safety analysis was performed with conservative tools. Two types of critics can be made. Firstly, conservative margins can be too large and it may be possible to reduce the cost of the plant or its operation with a best estimate approach. Secondly, some of the conservative hypotheses may notmore » really conservative in the full range of physical events which can occur during an accident. Simpson gives an interesting example: in some cases, the majoration of the residual power during a small break LOCA can lead to an overprediction of the swell level and thus of an overprediction of the core cooling, which is opposite to a conservative prediction. A last question is: does the accumulation of conservative hypotheses for a problem always give a conservative result? The two phase flow physics, mainly dealing with situation of mechanical and thermal non-equilibrium, is too much complicated to answer these questions with a simple engineer judgement. The objective of this paper is to make a review of the quantification of the uncertainties which can be made during code development and validation.« less

Spartan Release Engagement Mechanism (REM) stress and fracture analysis

NASA Technical Reports Server (NTRS)

Marlowe, D. S.; West, E. J.

1984-01-01

The revised stress and fracture analysis of the Spartan REM hardware for current load conditions and mass properties is presented. The stress analysis was performed using a NASTRAN math model of the Spartan REM adapter, base, and payload. Appendix A contains the material properties, loads, and stress analysis of the hardware. The computer output and model description are in Appendix B. Factors of safety used in the stress analysis were 1.4 on tested items and 2.0 on all other items. Fracture analysis of the items considered fracture critical was accomplished using the MSFC Crack Growth Analysis code. Loads and stresses were obtaind from the stress analysis. The fracture analysis notes are located in Appendix A and the computer output in Appendix B. All items analyzed met design and fracture criteria.

25 CFR 559.5 - What must a tribe submit to the Chairman with the copy of each facility license that has been...

Code of Federal Regulations, 2011 CFR

2011-04-01

... on the environment and the public health and safety, as well as mitigating measures the tribe may... environment and public health and safety. ... public health and safety laws, resolutions, codes, policies, standards or procedures applicable to its...

25 CFR 559.5 - What must a tribe submit to the Chairman with the copy of each facility license that has been...

Code of Federal Regulations, 2012 CFR

2012-04-01

... on the environment and the public health and safety, as well as mitigating measures the tribe may... environment and public health and safety. ... public health and safety laws, resolutions, codes, policies, standards or procedures applicable to its...

25 CFR 559.5 - What must a tribe submit to the Chairman with the copy of each facility license that has been...

Code of Federal Regulations, 2010 CFR

2010-04-01

... on the environment and the public health and safety, as well as mitigating measures the tribe may... environment and public health and safety. ... public health and safety laws, resolutions, codes, policies, standards or procedures applicable to its...

3 CFR 8672 - Proclamation 8672 of May 9, 2011. National Building Safety Month, 2011

Code of Federal Regulations, 2012 CFR

2012-01-01

... public and private sectors—to implement effective standards and codes that sustain safe and resilient structures. We need innovation and partnerships at all levels of society to develop transformative... Proclamation Building safety is a critical component of our homeland security, our personal and public safety...

Bleacher Safety: What Do We Look for? What Can We Do?

ERIC Educational Resources Information Center

IEA Environmental Consultant, 1999

1999-01-01

Discusses safety issues surrounding aging bleacher systems, highlighting the following three primary safety considerations: space between seats and footboards; guardrails; and the structural provisions of the 1997 Uniform Building Code. Tips for bleacher accident-prevention assessment and excerpts from federal and Minnesota legislation on bleacher…

30 CFR 75.1107-16 - Inspection of fire suppression devices.

Code of Federal Regulations, 2011 CFR

2011-07-01

...-16 Section 75.1107-16 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Fire Protection Fire... Systems” (NFPA No. 11A—1970). National Fire Code No. 13A “Care and Maintenance of Sprinkler Systems” (NFPA...

30 CFR 75.1107-16 - Inspection of fire suppression devices.

Code of Federal Regulations, 2010 CFR

2010-07-01

...-16 Section 75.1107-16 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Fire Protection Fire... Systems” (NFPA No. 11A—1970). National Fire Code No. 13A “Care and Maintenance of Sprinkler Systems” (NFPA...

77 FR 54836 - Federal Motor Vehicle Safety Standards

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-06

... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration 49 CFR Part 571 Federal Motor Vehicle Safety Standards CFR Correction 0 In Title 49 of the Code of Federal Regulations... read as follows: Sec. 571.119 Standard No. 119; New pneumatic tires for motor vehicles with a GVWR of...

Uniform emergency codes: will they improve safety?

PubMed

2005-01-01

There are pros and cons to uniform code systems, according to emergency medicine experts. Uniformity can be a benefit when ED nurses and other staff work at several facilities. It's critical that your staff understand not only what the codes stand for, but what they must do when codes are called. If your state institutes a new system, be sure to hold regular drills to familiarize your ED staff.

Current and anticipated uses of the thermal hydraulics codes at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support thesemore » needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.« less

Genuine worker participation-an indispensable key to effective global OHS.

PubMed

Brown, Garrett

2009-01-01

Working conditions, including workplace safety, in global supply chains of products sold by transnational corporations have only marginally improved over the last 15 years despite the development of hundreds of corporate "codes of conduct," code monitoring systems, and an elaborate new "corporate social responsibility" industry. The two underlying reasons for the lack of significant change are: 1) a schizophrenic business model which fatally undermines "socially responsible" sourcing programs with unyielding dictates for the lowest possible production costs; and 2) the lack of any meaningful participation by shop-floor workers in plant safety programs. Only when trained, empowered, and active workers are an integral part of workplace safety programs will conditions improve over the long term.

Health Information Technology in Healthcare Quality and Patient Safety: Literature Review.

PubMed

Feldman, Sue S; Buchalter, Scott; Hayes, Leslie W

2018-06-04

The area of healthcare quality and patient safety is starting to use health information technology to prevent reportable events, identify them before they become issues, and act on events that are thought to be unavoidable. As healthcare organizations begin to explore the use of health information technology in this realm, it is often unclear where fiscal and human efforts should be focused. The purpose of this study was to provide a foundation for understanding where to focus health information technology fiscal and human resources as well as expectations for the use of health information technology in healthcare quality and patient safety. A literature review was conducted to identify peer-reviewed publications reporting on the actual use of health information technology in healthcare quality and patient safety. Inductive thematic analysis with open coding was used to categorize a total of 41 studies. Three pre-set categories were used: prevention, identification, and action. Three additional categories were formed through coding: challenges, outcomes, and location. This study identifies five main categories across seven study settings. A majority of the studies used health IT for identification and prevention of healthcare quality and patient safety issues. In this realm, alerts, clinical decision support, and customized health IT solutions were most often implemented. Implementation, interface design, and culture were most often noted as challenges. This study provides valuable information as organizations determine where they stand to get the most "bang for their buck" relative to health IT for quality and patient safety. Knowing what implementations are being effectivity used by other organizations helps with fiscal and human resource planning as well as managing expectations relative to cost, scope, and outcomes. The findings from this scan of the literature suggest that having organizational champion leaders that can shepherd implementation, impact culture, and bridge knowledge with developers would be a valuable resource allocation to consider. ©Sue S Feldman, Scott Buchalter, Leslie W. Hayes. Originally published in JMIR Medical Informatics (http://medinform.jmir.org), 04.06.2018.

MODFLOW 2.0: A program for predicting moderator flow patterns

NASA Astrophysics Data System (ADS)

Peterson, P. F.; Paik, I. K.

1991-07-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in the operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors.

Health and Safety in Family Day Care Homes: Association Between Regulatory Non-compliance and Lower Median Income.

PubMed

Rosenthal, Marjorie S; Jeon, Sangchoon; Crowley, Angela A

2016-05-01

To determine frequency of non-compliance with child care regulations among family day care homes (FDCH) and identify the role of income in compliance. We analyzed non-compliance in 746 routine, unannounced inspection and re-inspection reports of FDCH collected by the Connecticut Department of Public Health licensing specialists in 2007-2008 and linked results to median income of zip code data. We grouped the 83 state regulations into 12 regulation categories, analyzed 11 categories, and used latent class analysis to classify each FDCH as high or low compliance for each category. We used logistic regression analysis to estimate the odds ratios of low compliance. Among the 746 FDCH inspections (594 first inspections and 152 re-inspections), we found high rates of non-compliance in inspection regulations in immunizations (32.9 %), water temperature (35.6 %) and hazards (30.0 %). Among the 11 regulation categories, 4 categories (indoor safety, emergency preparedness, child/family/staff documentation, and qualifications of provider) had regulations with high non-compliance. Median household income of FDCH zip code was lower for re-inspection sites than for inspection sites ($34,715 vs. $57,118, p < 0.0001) and FDCH in the lowest quartile of income had greater odds of low compliance in indoor safety (OR 1.86, 95 % CI 1.04, 3.35, p < 0.05). The majority of FDCH were in compliance with the majority of regulations, yet there are glaring non-compliance issues in inspections and re-inspections and there are income-based inequities that place children at higher risk who are already at high risk for suboptimal health outcomes.

Review of the probabilistic failure analysis methodology and other probabilistic approaches for application in aerospace structural design

NASA Technical Reports Server (NTRS)

Townsend, J.; Meyers, C.; Ortega, R.; Peck, J.; Rheinfurth, M.; Weinstock, B.

1993-01-01

Probabilistic structural analyses and design methods are steadily gaining acceptance within the aerospace industry. The safety factor approach to design has long been the industry standard, and it is believed by many to be overly conservative and thus, costly. A probabilistic approach to design may offer substantial cost savings. This report summarizes several probabilistic approaches: the probabilistic failure analysis (PFA) methodology developed by Jet Propulsion Laboratory, fast probability integration (FPI) methods, the NESSUS finite element code, and response surface methods. Example problems are provided to help identify the advantages and disadvantages of each method.

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Building Code Compliance and Enforcement: The Experience of SanFrancisco's Residential Energy Conservation Ordinanace and California'sBuildign Standards for New Construction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vine, E.

1990-11-01

As part of Lawrence Berkeley Laboratory's (LBL) technical assistance to the Sustainable City Project, compliance and enforcement activities related to local and state building codes for existing and new construction were evaluated in two case studies. The analysis of the City of San Francisco's Residential Energy Conservation Ordinance (RECO) showed that a limited, prescriptive energy conservation ordinance for existing residential construction can be enforced relatively easily with little administrative costs, and that compliance with such ordinances can be quite high. Compliance with the code was facilitated by extensive publicity, an informed public concerned with the cost of energy and knowledgeablemore » about energy efficiency, the threat of punishment (Order of Abatement), the use of private inspectors, and training workshops for City and private inspectors. The analysis of California's Title 24 Standards for new residential and commercial construction showed that enforcement of this type of code for many climate zones is more complex and requires extensive administrative support for education and training of inspectors, architects, engineers, and builders. Under this code, prescriptive and performance approaches for compliance are permitted, resulting in the demand for alternative methods of enforcement: technical assistance, plan review, field inspection, and computer analysis. In contrast to existing construction, building design and new materials and construction practices are of critical importance in new construction, creating a need for extensive technical assistance and extensive interaction between enforcement personnel and the building community. Compliance problems associated with building design and installation did occur in both residential and nonresidential buildings. Because statewide codes are enforced by local officials, these problems may increase over time as energy standards change and become more complex and as other standards (eg, health and safety codes) remain a higher priority. The California Energy Commission realizes that code enforcement by itself is insufficient and expects that additional educational and technical assistance efforts (eg, manuals, training programs, and toll-free telephone lines) will ameliorate these problems.« less

[Occupational medicine: practice and ethical requirements of the new law on health and safety in the workplace (legislative decree 81/2008)].

PubMed

Franco, Giuliano; Mora, Erika

2009-01-01

Decisions in occupational health may involve ethical conflicts arising from conflicts between stakeholders' interests. Codes of ethics can provide a practical guide to solve dilemmas. The new law on health and safety in the workplace in Italy (decree 81/2008) states that occupational health practice must comply with the code of ethics of the International Commission on Occupational Health. The universally acknowledged ethical principles of beneficience/nonmaleficience, autonomy and justice, which are the basis of the Charter of fundamental rights of the European Union, inspired this code. Although the code is not a systematic textbook of occupational health ethics and does not cover all possible aspects arising from the practice, making decisions based on it will assure their effectiveness and compliance with ethical principles, besides the formal respect of the law.

School Dress Codes and Uniform Policies.

ERIC Educational Resources Information Center

Anderson, Wendell

2002-01-01

Opinions abound on what students should wear to class. Some see student dress as a safety issue; others see it as a student-rights issue. The issue of dress codes and uniform policies has been tackled in the classroom, the boardroom, and the courtroom. This Policy Report examines the whole fabric of the debate on dress codes and uniform policies…

Rural School District Dress Code Implementation: Perceptions of Stakeholders after First Year

ERIC Educational Resources Information Center

Wright, Krystal M.

2012-01-01

Schools are continuously searching for solutions to solve truancy, academic, behavioral, safety, and climate issues. One of the latest trends in education is requiring students to adhere to dress codes as a solution to these issues. Dress codes can range from slightly restrictive clothing to the requiring of a uniform. Many school district…

Examination of adult and child bicyclist safety-relevant events using naturalistic bicycling methodology.

PubMed

Hamann, Cara J; Peek-Asa, Corinne

2017-05-01

Among roadway users, bicyclists are considered vulnerable due to their high risk for injury when involved in a crash. Little is known about the circumstances leading to near crashes, crashes, and related injuries or how these vary by age and gender. The purpose of this study was to examine the rates and characteristics of safety-relevant events (crashes, near crashes, errors, and traffic violations) among adult and child bicyclists. Bicyclist trips were captured using Pedal Portal, a data acquisition and coding system which includes a GPS-enabled video camera and graphical user interface. A total of 179 safety-relevant events were manually coded from trip videos. Overall, child errors and traffic violations occurred at a rate of 1.9 per 100min of riding, compared to 6.3 for adults. However, children rode on the sidewalk 56.4% of the time, compared with 12.7% for adults. For both adults and children, the highest safety-relevant event rates occurred on paved roadways with no bicycle facilities present (Adults=8.6 and Children=7.2, per 100min of riding). Our study, the first naturalistic study to compare safety-relevant events among adults and children, indicates large variation in riding behavior and exposure between child and adult bicyclists. The majority of identified events were traffic violations and we were not able to code all risk-relevant data (e.g., subtle avoidance behaviors, failure to check for traffic, probability of collision). Future naturalistic cycling studies would benefit from enhanced instrumentation (e.g., additional camera views) and coding protocols able to fill these gaps. Copyright © 2017 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, James; Goins, Monty; Paul, Pran

This safety analysis report for packaging (SARP) presents the results of the safety analysis prepared in support of the Consolidated Nuclear Security, LLC (CNS) request for licensing of the Model ES-3100 package with bulk highly enriched uranium (HEU) contents and issuance of a Type B(U) Fissile Material Certificate of Compliance. This SARP, published in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guide 7.9 and using information provided in UCID-21218 and NRC Regulatory Guide 7.10, demonstrates that the Y-12 National Security Complex (Y-12) ES-3100 package with bulk HEU contents meets the established NRC regulations for packaging, preparation formore » shipment, and transportation of radioactive materials given in Title 10, Part 71, of the Code of Federal Regulations (CFR) [10 CFR 71] as well as U.S. Department of Transportation (DOT) regulations for packaging and shipment of hazardous materials given in Title 49 CFR. To protect the health and safety of the public, shipments of adioactive materials are made in packaging that is designed, fabricated, assembled, tested, procured, used, maintained, and repaired in accordance with the provisions cited above. Safety requirements addressed by the regulations that must be met when transporting radioactive materials are containment of radioactive materials, radiation shielding, and assurance of nuclear subcriticality.« less

A Preliminary Assessment of the SURF Reactive Burn Model Implementation in FLAG

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Carl Edward; McCombe, Ryan Patrick; Carver, Kyle

Properly validated and calibrated reactive burn models (RBM) can be useful engineering tools for assessing high explosive performance and safety. Experiments with high explosives are expensive. Inexpensive RBM calculations are increasingly relied on for predictive analysis for performance and safety. This report discusses the validation of Menikoff and Shaw’s SURF reactive burn model, which has recently been implemented in the FLAG code. The LANL Gapstick experiment is discussed as is its’ utility in reactive burn model validation. Data obtained from pRad for the LT-63 series is also presented along with FLAG simulations using SURF for both PBX 9501 and PBXmore » 9502. Calibration parameters for both explosives are presented.« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE PAGES

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen; ...

2017-05-23

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

FDA adverse Event Problem Codes: standardizing the classification of device and patient problems associated with medical device use.

PubMed

Reed, Terrie L; Kaufman-Rivi, Diana

2010-01-01

The broad array of medical devices and the potential for device failures, malfunctions, and other adverse events associated with each device creates a challenge for public health device surveillance programs. Coding reported events by type of device problem provides one method for identifying a potential signal of a larger device issue. The Food and Drug Administration's (FDA) Center for Devices and Radiological Health (CDRH) Event Problem Codes that are used to report adverse events previously lacked a structured set of controls for code development and maintenance. Over time this led to inconsistent, ambiguous, and duplicative concepts being added to the code set on an ad-hoc basis. Recognizing the limitation of its coding system the FDA set out to update the system to improve its usefulness within FDA and as a basis of a global standard to identify important patient and device outcomes throughout the medical community. In 2004, FDA and the National Cancer Institute (NCI) signed a Memorandum of Understanding (MOU) whereby NCI agreed to provide terminology development and maintenance services to all FDA Centers. Under this MOU, CDRH's Office of Surveillance and Biometrics (OSB) convened a cross-Center workgroup and collaborated with staff at NCI Enterprise Vocabulary Service (EVS) to streamline the Patient and Device Problem Codes and integrate them into the NCI Thesaurus and Meta-Thesaurus. This initiative included many enhancements to the Event Problem Codes aimed at improving code selection as well as improving adverse event report analysis. LIMITATIONS & RECOMMENDATIONS: Staff resources, database concerns, and limited collaboration with external groups in the initial phases of the project are discussed. Adverse events associated with medical device use can be better understood when they are reported using a consistent and well-defined code set. This FDA initiative was an attempt to improve the structure and add control mechanisms to an existing code set, improve analysis tools that will better identify device safety trends, and improve the ability to prevent or mitigate effects of adverse events associated with medical device use.

Improving the safety of street-vended food.

PubMed

Moy, G; Hazzard, A; Käferstein, F

1997-01-01

An integrated plan of action for improving street food involving health and other regulatory authorities, vendors and consumers should address not only food safety, but also environmental health management, including consideration of inadequate sanitation and waste management, possible environmental pollution, congestion and disturbances to traffic. However, WHO cautions that, in view of their importance in the diets of urban populations, particularly the socially disadvantaged, every effort should be made to preserve the benefits provided by varied, inexpensive and often nutritious street food. Therefore, authorities concerned with street food management must balance efforts aimed at reducing the negative aspects on the environment with the benefits of street food and its important role in the community. Health authorities charged with responsibility for food safety control should match risk management action to the level of assessed risk. The rigorous application of codes and enforcement of regulations more suited to larger and permanent food service establishments is unlikely to be justifiable. Such rigorous application of codes and regulations may result in disappearance of the trade with consequent aggravation of hunger and malnutrition. Moreover, most codes and regulations have not been based on any systematic identification and assessment of health hazards associated with different types of foods and operations as embodied in the HACCP approach which has been recognized by Codex as the most cost-effective means for promoting food safety. WHO encourages the development of regulations that empower vendors to take greater responsibility for the preparation of safe food, and of codes of practice based on the HACCP system.

75 FR 33696 - Safety Zone: July Firework Display in Captain of the Port, Puget Sound AOR

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-15

...-AA00 Safety Zone: July Firework Display in Captain of the Port, Puget Sound AOR AGENCY: Coast Guard... Captain of the Port, Puget Sound AOR. (a) Safety Zone. The following area is a designated safety zone: all..., Captain of the Port, Puget Sound. [FR Doc. 2010-14294 Filed 6-14-10; 8:45 am] BILLING CODE 9110-04-P ...

Case-crossover analysis of heat-coded deaths and vulnerable subpopulations: Oklahoma, 1990-2011

NASA Astrophysics Data System (ADS)

Moore, Brianna F.; Brooke Anderson, G.; Johnson, Matthew G.; Brown, Sheryll; Bradley, Kristy K.; Magzamen, Sheryl

2017-11-01

The extent of the association between temperature and heat-coded deaths, for which heat is the primary cause of death, remains largely unknown. We explored the association between temperature and heat-coded deaths and potential interactions with various demographic and environmental factors. A total of 335 heat-coded deaths that occurred in Oklahoma from 1990 through 2011 were identified using heat-related International Classification of Diseases codes, cause-of-death nomenclature, and narrative descriptions. Conditional logistic regression models examined the association between temperature and heat index on heat-coded deaths. Interaction by demographic factors (age, sex, marital status, living alone, outdoor/heavy labor occupations) and environmental factors (ozone, PM10, PM2.5) was also explored. Temperatures ≥99 °F (the median value) were associated with approximately five times higher odds of a heat-coded death as compared to temperatures <99 °F (adjusted OR = 4.9, 95% CI 3.3, 7.2). The effect estimates were attenuated when exposure to heat was characterized by heat index. The interaction results suggest that effect of temperature on heat-coded deaths may depend on sex and occupation. For example, the odds of a heat-coded death among outdoor/heavy labor workers exposed to temperatures ≥99 °F was greater than expected based on the sum of the individual effects (observed OR = 14.0, 95% CI 2.7, 72.0; expected OR = 4.1 [2.8 + 2.3-1.0]). Our results highlight the extent of the association between temperature and heat-coded deaths and emphasize the need for a comprehensive, multisource definition of heat-coded deaths. Furthermore, based on the interaction results, we recommend that states implement or expand heat safety programs to protect vulnerable subpopulations, such as outdoor workers.

Revenue sources for financing transportation safety activities in Virginia : phase three, feasibility of a surcharge on traffic fines.

DOT National Transportation Integrated Search

1981-01-01

Senate Bill 85, an action of the 1978 General Assembly, amended the Code of Virginia to provide, in part, that the Division of Highway Safety be succeeded by the newly created Department of Transportation Safety effective July 1, 1978. In its Declara...

78 FR 11092 - Safety and Health Regulations for Construction

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-15

... LABOR DEPARTMENT Occupational Safety and Health Administration 29 CFR Part 1926 Safety and Health Regulations for Construction CFR Correction In Title 29 of the Code of Federal Regulations, Part 1926, revised as of July 1, 2012, on page 225, in Sec. 1926.152, paragraph (c)(16) is added to read as follows: Sec...

75 FR 33162 - Airworthiness Directives; Microturbo Saphir 20 Model 095 Auxiliary Power Units (APUs)

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-11

... information (MCAI) issued by the European Aviation Safety Agency (EASA) to identify and correct an unsafe... States Code specifies the FAA's authority to issue rules on aviation safety. Subtitle I, section 106... the AD docket. List of Subjects in 14 CFR Part 39 Air transportation, Aircraft, Aviation safety...

Coupling Legacy and Contemporary Deterministic Codes to Goldsim for Probabilistic Assessments of Potential Low-Level Waste Repository Sites

NASA Astrophysics Data System (ADS)

Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.

2006-12-01

Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer project between Sandia National Laboratories and Taiwan's Institute of Nuclear Energy Research (INER) for the preliminary assessment of several candidate low-level waste repository sites. Sandia National Laboratories is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under Contract DE AC04 94AL85000.

A Theory of False Cognitive Expectancies in Airline Pilots

NASA Astrophysics Data System (ADS)

Cortes, Antonio I.

The Theory of False Cognitive Expectancies was developed by studying high reliability flight operations. Airline pilots depend extensively on cognitive expectancies to perceive, understand, and predict actions and events. Out of 1,363 incident reports submitted by airline pilots to the National Aeronautics and Space Administration Aviation Safety Reporting System over a year's time, 110 reports were found to contain evidence of 127 false cognitive expectancies in pilots. A comprehensive taxonomy was developed with six categories of interest. The dataset of 127 false expectancies was used to initially code tentative taxon values for each category. Intermediate coding through constant comparative analysis completed the taxonomy. The taxonomy was used for the advanced coding of chronological context-dependent visualizations of expectancy factors, known as strands, which depict the major factors in the creation and propagation of each expectancy. Strands were mapped into common networks to detect highly represented expectancy processes. Theoretical integration established 11 sources of false expectancies, the most common expectancy errors, and those conspicuous factors worthy of future study. The most prevalent source of false cognitive expectancies within the dataset was determined to be unconscious individual modeling based on past events. Integrative analyses also revealed relationships between expectancies and flight deck automation, unresolved discrepancies, and levels of situation awareness. Particularly noteworthy were the findings that false expectancies can combine in three possible permutations to diminish situation awareness and examples of how false expectancies can be unwittingly transmitted from one person to another. The theory resulting from this research can enhance the error coding process used during aircraft line oriented safety audits, lays the foundation for developing expectancy management training programs, and will allow researchers to proffer hypotheses for human testing using flight simulators.

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

NASA Astrophysics Data System (ADS)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

Ensuring the validity of calculated subcritical limits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionallymore » subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin.« less

Certifying Domain-Specific Policies

NASA Technical Reports Server (NTRS)

Lowry, Michael; Pressburger, Thomas; Rosu, Grigore; Koga, Dennis (Technical Monitor)

2001-01-01

Proof-checking code for compliance to safety policies potentially enables a product-oriented approach to certain aspects of software certification. To date, previous research has focused on generic, low-level programming-language properties such as memory type safety. In this paper we consider proof-checking higher-level domain -specific properties for compliance to safety policies. The paper first describes a framework related to abstract interpretation in which compliance to a class of certification policies can be efficiently calculated Membership equational logic is shown to provide a rich logic for carrying out such calculations, including partiality, for certification. The architecture for a domain-specific certifier is described, followed by an implemented case study. The case study considers consistency of abstract variable attributes in code that performs geometric calculations in Aerospace systems.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Estimation of the behavior factor of existing RC-MRF buildings

NASA Astrophysics Data System (ADS)

Vona, Marco; Mastroberti, Monica

2018-01-01

In recent years, several research groups have studied a new generation of analysis methods for seismic response assessment of existing buildings. Nevertheless, many important developments are still needed in order to define more reliable and effective assessment procedures. Moreover, regarding existing buildings, it should be highlighted that due to the low knowledge level, the linear elastic analysis is the only analysis method allowed. The same codes (such as NTC2008, EC8) consider the linear dynamic analysis with behavior factor as the reference method for the evaluation of seismic demand. This type of analysis is based on a linear-elastic structural model subject to a design spectrum, obtained by reducing the elastic spectrum through a behavior factor. The behavior factor (reduction factor or q factor in some codes) is used to reduce the elastic spectrum ordinate or the forces obtained from a linear analysis in order to take into account the non-linear structural capacities. The behavior factors should be defined based on several parameters that influence the seismic nonlinear capacity, such as mechanical materials characteristics, structural system, irregularity and design procedures. In practical applications, there is still an evident lack of detailed rules and accurate behavior factor values adequate for existing buildings. In this work, some investigations of the seismic capacity of the main existing RC-MRF building types have been carried out. In order to make a correct evaluation of the seismic force demand, actual behavior factor values coherent with force based seismic safety assessment procedure have been proposed and compared with the values reported in the Italian seismic code, NTC08.

Feasibility Study of Alternative Fabrication Methods.

DTIC Science & Technology

1979-08-01

must comply with the I requirements of the latest edition of the National Electrical Code. The Body and Liner Assembly System will comply with I the...latest edition of the National Electrical Code per AMCR 385 (Army Material Command Safety Manual). Also, OSHA’s 1 Occupational Safety and Health...the top of the elevator. On the top and at the rear of * A-4 50 UD( AISA 6 --7 G OVe CASE THOMS ’ON J .7O37IA (PLACES) _ _ -- 3. BAL - 1N-TOS10 12N

78 FR 77773 - Shipping Coordinating Committee; Notice of Committee Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-24

... the International Code of safety for ships using gases or low flash-point fuels (IGF Code) Members of..., by fax at (202) 372-8283, or in writing at Commandant (CG-OES-1), U.S. Coast Guard Stop 7509, 2703...

Methodology, status, and plans for development and assessment of the RELAP5 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, G.W.; Riemke, R.A.

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

Element Load Data Processor (ELDAP) Users Manual

NASA Technical Reports Server (NTRS)

Ramsey, John K., Jr.; Ramsey, John K., Sr.

2015-01-01

Often, the shear and tensile forces and moments are extracted from finite element analyses to be used in off-line calculations for evaluating the integrity of structural connections involving bolts, rivets, and welds. Usually the maximum forces and moments are desired for use in the calculations. In situations where there are numerous structural connections of interest for numerous load cases, the effort in finding the true maximum force and/or moment combinations among all fasteners and welds and load cases becomes difficult. The Element Load Data Processor (ELDAP) software described herein makes this effort manageable. This software eliminates the possibility of overlooking the worst-case forces and moments that could result in erroneous positive margins of safety and/or selecting inconsistent combinations of forces and moments resulting in false negative margins of safety. In addition to forces and moments, any scalar quantity output in a PATRAN report file may be evaluated with this software. This software was originally written to fill an urgent need during the structural analysis of the Ares I-X Interstage segment. As such, this software was coded in a straightforward manner with no effort made to optimize or minimize code or to develop a graphical user interface.

Semantic Annotation of Complex Text Structures in Problem Reports

NASA Technical Reports Server (NTRS)

Malin, Jane T.; Throop, David R.; Fleming, Land D.

2011-01-01

Text analysis is important for effective information retrieval from databases where the critical information is embedded in text fields. Aerospace safety depends on effective retrieval of relevant and related problem reports for the purpose of trend analysis. The complex text syntax in problem descriptions has limited statistical text mining of problem reports. The presentation describes an intelligent tagging approach that applies syntactic and then semantic analysis to overcome this problem. The tags identify types of problems and equipment that are embedded in the text descriptions. The power of these tags is illustrated in a faceted searching and browsing interface for problem report trending that combines automatically generated tags with database code fields and temporal information.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Next-Generation Telemetry Workstation

NASA Technical Reports Server (NTRS)

2008-01-01

A next-generation telemetry workstation has been developed to replace the one currently used to test and control Range Safety systems. Improving upon the performance of the original system, the new telemetry workstation uses dual-channel telemetry boards for better synchronization of the two uplink telemetry streams. The new workstation also includes an Interrange Instrumentation Group/Global Positioning System (IRIG/GPS) time code receiver board for independent, local time stamping of return-link data. The next-generation system will also record and play back return-link data for postlaunch analysis.

DOE Hydrogen Program: 2006 Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milliken, J.

This report summarizes comments from the Peer Review Panel at the FY 2006 DOE Hydrogen Program Annual Merit Review, held on May 16-19, 2006, in Arlington, Virginia. The projects evaluated support the Department of Energy and President Bush's Hydrogen Initiative. The results of this merit review and peer evaluation are major inputs used by DOE to make funding decisions. Project areas include hydrogen production and delivery; hydrogen storage; fuel cells; technology validation; safety, codes and standards; education; and systems analysis.

DOE Hydrogen Program: 2005 Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chalk, S. G.

This report summarizes comments from the Peer Review Panel at the FY 2005 DOE Hydrogen Program Annual Merit Review, held on May 23-26, 2005, in Arlington, Virginia. The projects evaluated support the Department of Energy and President Bush's Hydrogen Initiative. The results of this merit review and peer evaluation are major inputs used by DOE to make funding decisions. Project areas include hydrogen production and delivery; hydrogen storage; fuel cells; technology validation; safety, codes and standards; education; and systems analysis.

DOE Hydrogen Program: 2007 Annual Merit Review and Peer Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milliken, J.

This report summarizes comments from the Peer Review Panel at the FY 2007 DOE Hydrogen Program Annual Merit Review, held on May 14-18, 2007, in Washington, D.C. The projects evaluated support the Department of Energy and President Bush's Hydrogen Initiative. The results of this merit review and peer evaluation are major inputs used by DOE to make funding decisions. Project areas include hydrogen production and delivery; hydrogen storage; fuel cells; technology validation; safety, codes and standards; education; and systems analysis.

Automated Translation of Safety Critical Application Software Specifications into PLC Ladder Logic

NASA Technical Reports Server (NTRS)

Leucht, Kurt W.; Semmel, Glenn S.

2008-01-01

The numerous benefits of automatic application code generation are widely accepted within the software engineering community. A few of these benefits include raising the abstraction level of application programming, shorter product development time, lower maintenance costs, and increased code quality and consistency. Surprisingly, code generation concepts have not yet found wide acceptance and use in the field of programmable logic controller (PLC) software development. Software engineers at the NASA Kennedy Space Center (KSC) recognized the need for PLC code generation while developing their new ground checkout and launch processing system. They developed a process and a prototype software tool that automatically translates a high-level representation or specification of safety critical application software into ladder logic that executes on a PLC. This process and tool are expected to increase the reliability of the PLC code over that which is written manually, and may even lower life-cycle costs and shorten the development schedule of the new control system at KSC. This paper examines the problem domain and discusses the process and software tool that were prototyped by the KSC software engineers.

Analyses in Support of Risk-Informed Natural Gas Vehicle Maintenance Facility Codes and Standards: Phase II.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaylock, Myra L.; LaFleur, Chris Bensdotter; Muna, Alice Baca

Safety standards development for maintenance facilities of liquid and compressed natural gas fueled vehicles is required to ensure proper facility design and operating procedures. Standard development organizations are utilizing risk-informed concepts to develop natural gas vehicle (NGV) codes and standards so that maintenance facilities meet acceptable risk levels. The present report summarizes Phase II work for existing NGV repair facility code requirements and highlights inconsistencies that need quantitative analysis into their effectiveness. A Hazardous and Operability study was performed to identify key scenarios of interest using risk ranking. Detailed simulations and modeling were performed to estimate the location and behaviormore » of natural gas releases based on these scenarios. Specific code conflicts were identified, and ineffective code requirements were highlighted and resolutions proposed. These include ventilation rate basis on area or volume, as well as a ceiling offset which seems ineffective at protecting against flammable gas concentrations. ACKNOWLEDGEMENTS The authors gratefully acknowledge Bill Houf (SNL -- Retired) for his assistance with the set-up and post-processing of the numerical simulations. The authors also acknowledge Doug Horne (retired) for his helpful discussions. We would also like to acknowledge the support from the Clean Cities program of DOE's Vehicle Technology Office.« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Static-stress analysis of dual-axis safety vessel

NASA Astrophysics Data System (ADS)

Bultman, D. H.

1992-11-01

An 8 ft diameter safety vessel, made of HSLA-100 steel, is evaluated to determine its ability to contain the quasi-static residual pressure from a high explosive (HE) blast. The safety vessel is designed for use with the Dual-Axis Radiographic Hydrotest (DARHT) facility being developed at Los Alamos National Laboratory. A smaller confinement vessel fits inside the safety vessel and contains the actual explosion, and the safety vessel functions as a second layer of containment in the unlikely case of a confinement vessel leak. The safety vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section 8, Division 1, and the Welding Research Council Bulletin, WRC107. Combined stresses that result from internal pressure and external loads on nozzles are calculated and compared to the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzle components are adequately designed for a static pressure of 830 psi, plus the maximum expected external loads. Shell stresses at the 'shell to nozzle' interface, produced from external loads on the nozzles, were less than 700 psi. The maximum combined stress resulting from the internal pressure plus external loads was 17,384 psi, which is significantly less than the allowable stress of 42,375 psi for HSLA-100 steel.

The pedestrian in the transportation system : proposed traffic safety legislation.

DOT National Transportation Integrated Search

1981-01-01

The purposes of this project were to review and evaluate Virginia's traffic laws related to pedestrians, compare provisions of the Code of Virginia with those of the statutes of other states and the Uniform Vehicle Code, and, if appropriate, propose ...

Deterministic Local Sensitivity Analysis of Augmented Systems - II: Applications to the QUENCH-04 Experiment Using the RELAP5/MOD3.2 Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ionescu-Bujor, Mihaela; Jin Xuezhou; Cacuci, Dan G.

2005-09-15

The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less

Preliminary Analysis of the BASALA-H Experimental Programme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaise, Patrick; Fougeras, Philippe; Philibert, Herve

2002-07-01

This paper is focused on the preliminary analysis of results obtained on the first cores of the first phase of the BASALA (Boiling water reactor Advanced core physics Study Aimed at mox fuel Lattice) programme, aimed at studying the neutronic parameters in ABWR core in hot conditions, currently under investigation in the French EOLE critical facility, within the framework of a cooperation between NUPEC, CEA and Cogema. The first 'on-line' analysis of the results has been made, using a new preliminary design and safety scheme based on the French APOLLO-2 code in its 2.4 qualified version and associated CEA-93 V4more » (JEF-2.2) Library, that will enable the Experimental Physics Division (SPEx) to perform future core designs. It describes the scheme adopted and the results obtained in various cases, going to the critical size determination to the reactivity worth of the perturbed configurations (voided, over-moderated, and poisoned with Gd{sub 2}O{sub 3}-UO{sub 2} pins). A preliminary study on the experimental results on the MISTRAL-4 is resumed, and the comparison of APOLLO-2 versus MCNP-4C calculations on these cores is made. The results obtained show very good agreements between the two codes, and versus the experiment. This work opens the way to the future full analysis of the experimental results of the qualifying teams with completely validated schemes, based on the new 2.5 version of the APOLLO-2 code. (authors)« less

Analysis of construction accidents in Spain, 2003-2008.

PubMed

López Arquillos, Antonio; Rubio Romero, Juan Carlos; Gibb, Alistair

2012-12-01

The research objective for this paper is to obtain a new extended and updated insight to the likely causes of construction accidents in Spain, in order to identify suitable mitigating actions. The paper analyzes all construction sector accidents in Spain between 2003 and 2008. Ten variables were chosen and the influence of each variable is evaluated with respect to the severity of the accident. The descriptive analysis is based on a total of 1,163,178 accidents. Results showed that the severity of accidents was related to variables including age, CNAE (National Classification of Economic Activities) code, size of company, length of service, location of accident, day of the week, days of absence, deviation, injury, and climatic zones. According to data analyzed, a large company is not always necessarily safer than a small company in the aspect of fatal accidents, experienced workers do not have the best accident fatality rates, and accidents occurring away from the usual workplace had more severe consequences. Results obtained in this paper can be used by companies in their occupational safety strategies, and in their safety training programs. Copyright © 2012 National Safety Council and Elsevier Ltd. All rights reserved.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

29 CFR 1926.402 - Applicability.

Code of Federal Regulations, 2014 CFR

2014-07-01

...) SAFETY AND HEALTH REGULATIONS FOR CONSTRUCTION Electrical Installation Safety Requirements § 1926.402... electrical equipment and installations used to provide electric power and light at the jobsite. These... commenced. Note: If the electrical installation is made in accordance with the National Electrical Code ANSI...

78 FR 37505 - International Standard-Setting Activities

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... Code of Practice for Weed Control to Prevent and Reduce Pyrrolizidine Alkaloid Contamination in Food... DEPARTMENT OF AGRICULTURE Food Safety and Inspection Service [Docket No. FSIS-2013-0002] International Standard-Setting Activities AGENCY: Office of Food Safety, USDA. ACTION: Notice. SUMMARY: This...

Mechanical code comparator

DOEpatents

Peter, Frank J.; Dalton, Larry J.; Plummer, David W.

2002-01-01

A new class of mechanical code comparators is described which have broad potential for application in safety, surety, and security applications. These devices can be implemented as micro-scale electromechanical systems that isolate a secure or otherwise controlled device until an access code is entered. This access code is converted into a series of mechanical inputs to the mechanical code comparator, which compares the access code to a pre-input combination, entered previously into the mechanical code comparator by an operator at the system security control point. These devices provide extremely high levels of robust security. Being totally mechanical in operation, an access control system properly based on such devices cannot be circumvented by software attack alone.

Evaluation of Key Factors Impacting Feeding Safety in the Neonatal Intensive Care Unit: A Systematic Review.

PubMed

Matus, Bethany A; Bridges, Kayla M; Logomarsino, John V

2018-06-21

Individualized feeding care plans and safe handling of milk (human or formula) are critical in promoting growth, immune function, and neurodevelopment in the preterm infant. Feeding errors and disruptions or limitations to feeding processes in the neonatal intensive care unit (NICU) are associated with negative safety events. Feeding errors include contamination of milk and delivery of incorrect or expired milk and may result in adverse gastrointestinal illnesses. The purpose of this review was to evaluate the effect(s) of centralized milk preparation, use of trained technicians, use of bar code-scanning software, and collaboration between registered dietitians and registered nurses on feeding safety in the NICU. A systematic review of the literature was completed, and 12 articles were selected as relevant to search criteria. Study quality was evaluated using the Downs and Black scoring tool. An evaluation of human studies indicated that the use of centralized milk preparation, trained technicians, bar code-scanning software, and possible registered dietitian involvement decreased feeding-associated error in the NICU. A state-of-the-art NICU includes a centralized milk preparation area staffed by trained technicians, care supported by bar code-scanning software, and utilization of a registered dietitian to improve patient safety. These resources will provide nurses more time to focus on nursing-specific neonatal care. Further research is needed to evaluate the impact of factors related to feeding safety in the NICU as well as potential financial benefits of these quality improvement opportunities.

Development of Safety Assessment Code for Decommissioning of Nuclear Facilities

NASA Astrophysics Data System (ADS)

Shimada, Taro; Ohshima, Soichiro; Sukegawa, Takenori

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (currently JAEA). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning, and also evaluates the public dose at accidental situations including fire and explosion. As for the public, both the internal and the external doses are calculated by considering inhalation, ingestion, direct radiation from radioactive aerosols and radioactive depositions, and skyshine radiation from waste containers. For external dose for workers, the dose rate from contaminated components and structures to be dismantled is calculated. Internal dose for workers is calculated by considering dismantling conditions, e.g. cutting speed, cutting length of the components and exhaust velocity. Estimation models for dose rate and staying time were verified by comparison with the actual external dose of workers which were acquired during JPDR decommissioning project. DecDose code is expected to contribute the safety assessment for decommissioning of nuclear facilities.

Commercial D-T FRC Power Plant Systems Analysis

NASA Astrophysics Data System (ADS)

Nguyen, Canh; Santarius, John; Emmert, Gilbert; Steinhauer, Loren; Stubna, Michael

1998-11-01

Results of an engineering issues scoping study of a Field-Reversed Configuration (FRC) burning D-T fuel will be presented. The study primarily focuses on engineering issues, such as tritium-breeding blanket design, radiation shielding, neutron damage, activation, safety, and environment. This presentation will concentrate on plasma physics, current drive, economics, and systems integration, which are important for the overall systems analysis. A systems code serves as the key tool in defining a reference point for detailed physics and engineering calculations plus parametric variations, and typical cases will be presented. Advantages of the cylindrical geometry and high beta (plasma pressure/magnetic-field pressure) are evident.

Bar Coding and Tracking in Pathology.

PubMed

Hanna, Matthew G; Pantanowitz, Liron

2016-03-01

Bar coding and specimen tracking are intricately linked to pathology workflow and efficiency. In the pathology laboratory, bar coding facilitates many laboratory practices, including specimen tracking, automation, and quality management. Data obtained from bar coding can be used to identify, locate, standardize, and audit specimens to achieve maximal laboratory efficiency and patient safety. Variables that need to be considered when implementing and maintaining a bar coding and tracking system include assets to be labeled, bar code symbologies, hardware, software, workflow, and laboratory and information technology infrastructure as well as interoperability with the laboratory information system. This article addresses these issues, primarily focusing on surgical pathology. Copyright © 2016 Elsevier Inc. All rights reserved.

Bar Coding and Tracking in Pathology.

PubMed

Hanna, Matthew G; Pantanowitz, Liron

2015-06-01

Bar coding and specimen tracking are intricately linked to pathology workflow and efficiency. In the pathology laboratory, bar coding facilitates many laboratory practices, including specimen tracking, automation, and quality management. Data obtained from bar coding can be used to identify, locate, standardize, and audit specimens to achieve maximal laboratory efficiency and patient safety. Variables that need to be considered when implementing and maintaining a bar coding and tracking system include assets to be labeled, bar code symbologies, hardware, software, workflow, and laboratory and information technology infrastructure as well as interoperability with the laboratory information system. This article addresses these issues, primarily focusing on surgical pathology. Copyright © 2015 Elsevier Inc. All rights reserved.

Long term integrity for space station power systems

NASA Technical Reports Server (NTRS)

Leckie, F. A.; Marriott, D. L.

1991-01-01

A study was made of the High Temperature Design Codes ASME N47, British R5, and the French RCC-MR Rules. It is concluded that all these codes provide a good basis of design for space application. The new British R5 is the most complete since it deals with the problem of defects. The ASME N47 was subjected longer to practical application and scrutiny. A draft code is introduced, and a proposed draft for high temperature design in which attempts were made to identify gaps and improvements is suggested. The design is limited by creep characteristics. In these circumstances, life is strongly affected by the selected value of the factor of safety. The factor of safety of primary loads adopted in the codes is 1.5. Maybe a lower value of 1.25 is permissible for use in space. Long term creep rupture data for HAYNES 188 is deficient and it is suggested that extrapolation methods be investigated.

BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

DOE PAGES

Mandelli, D.; Smith, C.; Riley, T.; ...

2016-01-01

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Watermore » Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Furthermore, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.« less

77 FR 61513 - Information Disclosure Under Section 6(b) of the Consumer Product Safety Act

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-10

... CONSUMER PRODUCT SAFETY COMMISSION 16 CFR Part 1101 Information Disclosure Under Section 6(b) of the Consumer Product Safety Act CFR Correction 0 In Title 16 of the Code of Federal Regulations, Part 1000 to End, revised as of January 1, 2012, on page 147, in Sec. 1101.25 (a) and (b), the words ``5...

75 FR 57659 - Airworthiness Directives Eurocopter France (ECF) Model SA-365N1, AS-365N2, AS 365 N3, EC 155B...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-22

... aviation authority of the European Aviation Safety Agency (EASA). The MCAI AD reports the separation and...: DOT/FAA Southwest Region, Gary Roach, ASW-111, Aviation Safety Engineer, Rotorcraft Directorate... Rulemaking Title 49 of the United States Code specifies the FAA's authority to issue rules on aviation safety...

RELAP5-3D Resolution of Known Restart/Backup Issues

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mesina, George L.; Anderson, Nolan A.

2014-12-01

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequentialmore » verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.« less

Access to properly fitting personal protective equipment for female construction workers.

PubMed

Onyebeke, Lynn C; Papazaharias, Demetrios M; Freund, Alice; Dropkin, Jonathan; McCann, Michael; Sanchez, Sadie H; Hashim, Dana; Meyer, John D; Lucchini, Roberto G; Zuckerman, Norman C

2016-11-01

Previous literature suggests that most personal protective equipment (PPE) for construction is designed for males and does not accommodate female anthropometry. We conducted a pilot study to identify whether female construction workers currently have adequate access to properly fitting PPE. Semi-structured focus group interviews were conducted with union female carpenters, laborers, and ironworkers. Researchers coded focus group transcriptions and extracted major themes using thematic framework analysis. Participants (n = 23) had a mean of 15.1 years of construction experience (range 3-34.5 years). A majority reported fit problems for many types of PPE (gloves, harnesses, safety vests, work boots, outerwear), generally noting that the equipment provided by contractors was too large. Other emergent themes included female workers purchasing their own PPE, exposure to various safety hazards from poorly fitted PPE, and perceived indifferent safety culture. Female construction workers continue to have difficulty accessing properly fitting PPE. Am. J. Ind. Med. 59:1032-1040, 2016. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

Capabilities overview of the MORET 5 Monte Carlo code

NASA Astrophysics Data System (ADS)

Cochet, B.; Jinaphanh, A.; Heulers, L.; Jacquet, O.

2014-06-01

The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations, describe the materials, define their own tallies in order to analyse the results. The MORET code has been initially designed to perform calculations for criticality safety assessments. New features has been introduced in the MORET 5 code to expand its use for reactor applications. This paper presents an overview of the MORET 5 code capabilities, going through the description of materials, the geometry modelling, the transport simulation and the definition of the outputs.

Eighteen years of recommendations to prevent industrial chemical incidents: results and lessons learned of the US Chemical Safety Board.

PubMed

Tinney, V A; Anenberg, S C; Kaszniak, M; Robinson, B

2016-10-01

The US Chemical Safety Board (CSB), a federal agency that investigates significant chemical incidents and hazards, is interested in determining the impact of the recommendations resulting from its investigations, and how to better more effective recommendations to prevent chemical incidents. This is a descriptive study of the US Chemical Safety Board's safety recommendations. The CSB coded and analysed its safety recommendations according to potential impact on reducing incidents, implementation status, purpose and recipient type. As of March 31, 2015, the CSB has issued 733 recommendations, 75% (548) of which are closed and 25% (185) of which remain open. For recommendations categorised as having high, medium, and low impact, 38% (78), 76% (160), and 78% (245) were implemented, respectively. CSB recommendations have led to important and lasting safety changes through regulations, industry guidance and voluntary consensus standards, and individual companies; however, coding recommendations by potential impact do not fully capture the influence of CSB recommendations. While this methodology serves as a preliminary way to determine the effect of recommendations, further data are needed to determine the extent to which these safety changes have reduced the frequency or severity of industrial accidents. Copyright © 2016 The Royal Society for Public Health. All rights reserved.

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

Cheng, Lap-Yan; Wei, Thomas Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

Why is Probabilistic Seismic Hazard Analysis (PSHA) still used?

NASA Astrophysics Data System (ADS)

Mulargia, Francesco; Stark, Philip B.; Geller, Robert J.

2017-03-01

Even though it has never been validated by objective testing, Probabilistic Seismic Hazard Analysis (PSHA) has been widely used for almost 50 years by governments and industry in applications with lives and property hanging in the balance, such as deciding safety criteria for nuclear power plants, making official national hazard maps, developing building code requirements, and determining earthquake insurance rates. PSHA rests on assumptions now known to conflict with earthquake physics; many damaging earthquakes, including the 1988 Spitak, Armenia, event and the 2011 Tohoku, Japan, event, have occurred in regions relatively rated low-risk by PSHA hazard maps. No extant method, including PSHA, produces reliable estimates of seismic hazard. Earthquake hazard mitigation should be recognized to be inherently political, involving a tradeoff between uncertain costs and uncertain risks. Earthquake scientists, engineers, and risk managers can make important contributions to the hard problem of allocating limited resources wisely, but government officials and stakeholders must take responsibility for the risks of accidents due to natural events that exceed the adopted safety criteria.

Video capture of clinical care to enhance patient safety

PubMed Central

Weinger, M; Gonzales, D; Slagle, J; Syeed, M

2004-01-01



 Experience from other domains suggests that videotaping and analyzing actual clinical care can provide valuable insights for enhancing patient safety through improvements in the process of care. Methods are described for the videotaping and analysis of clinical care using a high quality portable multi-angle digital video system that enables simultaneous capture of vital signs and time code synchronization of all data streams. An observer can conduct clinician performance assessment (such as workload measurements or behavioral task analysis) either in real time (during videotaping) or while viewing previously recorded videotapes. Supplemental data are synchronized with the video record and stored electronically in a hierarchical database. The video records are transferred to DVD, resulting in a small, cheap, and accessible archive. A number of technical and logistical issues are discussed, including consent of patients and clinicians, maintaining subject privacy and confidentiality, and data security. Using anesthesiology as a test environment, over 270 clinical cases (872 hours) have been successfully videotaped and processed using the system. PMID:15069222

Safety engineering: KTA code of practice. Lifting mechanisms in nuclear plant

NASA Astrophysics Data System (ADS)

Lifting mechanisms safety requirements are discussed in accordance with the present state of development of science and engineering for the protection of life, health, and assets against the dangers of nuclear energy and the ill effects of ionizing radiation.

Preschool Comes to School: Design Concerns of Preschool Facilities.

ERIC Educational Resources Information Center

Passantino, Richard J.

1994-01-01

Addresses design and safety issues of which school business administrators should be aware when they become involved in integrating a preschool facility with an elementary school. Discusses building environmental factors, safety, health codes, play surfaces, energy conservation, and architectural considerations. (KDP)

Trends in readmission rates for safety net hospitals and non-safety net hospitals in the era of the US Hospital Readmission Reduction Program: a retrospective time series analysis using Medicare administrative claims data from 2008 to 2015.

PubMed

Salerno, Amy M; Horwitz, Leora I; Kwon, Ji Young; Herrin, Jeph; Grady, Jacqueline N; Lin, Zhenqiu; Ross, Joseph S; Bernheim, Susannah M

2017-07-13

To compare trends in readmission rates among safety net and non-safety net hospitals under the US Hospital Readmission Reduction Program (HRRP). A retrospective time series analysis using Medicare administrative claims data from January 2008 to June 2015. We examined 3254 US hospitals eligible for penalties under the HRRP, categorised as safety net or non-safety net hospitals based on the hospital's proportion of patients with low socioeconomic status. Admissions for Medicare fee-for-service patients, age ≥65 years, discharged alive, who had a valid five-digit zip code and did not have a principal discharge diagnosis of cancer or psychiatric illness were included, for a total of 52 516 213 index admissions. Mean hospital-level, all-condition, 30-day risk-adjusted standardised unplanned readmission rate, measured quarterly, along with quarterly rate of change, and an interrupted time series examining: April-June 2010, after HRRP was passed, and October-December 2012, after HRRP penalties were implemented. 58.0% (SD 15.3) of safety net hospitals and 17.1% (SD 10.4) of non-safety net hospitals' patients were in the lowest quartile of socioeconomic status. The mean safety net hospital standardised readmission rate declined from 17.0% (SD 3.7) to 13.6% (SD 3.6), whereas the mean non-safety net hospital declined from 15.4% (SD 3.0) to 12.7% (SD 2.5). The absolute difference in rates between safety net and non-safety net hospitals declined from 1.6% (95% CI 1.3 to 1.9) to 0.9% (0.7 to 1.2). The quarterly decline in standardised readmission rates was 0.03 percentage points (95% CI 0.03 to 0.02, p<0.001) greater among safety net hospitals over the entire study period, and no differential change among safety net and non-safety net hospitals was found after either HRRP was passed or penalties enacted. Since HRRP was passed and penalties implemented, readmission rates for safety net hospitals have decreased more rapidly than those for non-safety net hospitals. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

DOE PAGES

Perfetti, Christopher M.; Rearden, Bradley T.; Martin, William R.

2016-02-25

Sensitivity coefficients describe the fractional change in a system response that is induced by changes to system parameters and nuclear data. The Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the developmentmore » of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE-KENO framework of the SCALE code system to enable TSUNAMI-3D to perform eigenvalue sensitivity calculations using continuous-energy Monte Carlo methods. This work provides a detailed description of the theory behind the CLUTCH method and describes in detail its implementation. This work explores the improvements in eigenvalue sensitivity coefficient accuracy that can be gained through the use of continuous-energy sensitivity methods and also compares several sensitivity methods in terms of computational efficiency and memory requirements.« less

Pathways to Recurrent Trauma Among Young Black Men: Traumatic Stress, Substance Use, and the “Code of the Street”

PubMed Central

Rich, John A.; Grey, Courtney M.

2005-01-01

Recurrent interpersonal violence is a major cause of death and disability among young Black men. Quantitative studies have uncovered factors associated with reinjury, but little is known about how these factors work together. We interviewed young Black male victims to understand their experience of violence. Qualitative analysis of their narratives revealed how their struggle to reestablish safety shaped their response to injury. Aspects of the “code of the street” (including the need for respect) and lack of faith in the police combined with traumatic stress and substance use to accentuate their sense of vulnerability. Victims then reacted to protect themselves in ways that could increase their risk of reinjury. We describe a model with implications for reducing rates of recurrent violent injuries. PMID:15855457

Coupling the System Analysis Module with SAS4A/SASSYS-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Hu, R.

2016-09-30

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format formore » PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly attributed to the limited maturity of the SAM development effort. The severe accident modeling capabilities in SAS4A/SASSYS-1 (sodium boiling, fuel melting and relocation) will continue to play a vital role for a long time. Therefore, the SAS4A/SASSYS-1 modernization effort should remain a high priority task under the ART program to ensure continued participation in domestic and international SFR safety collaborations and design optimizations. On the other hand, SAM provides an advanced system analysis tool, with improved numerical solution schemes, data management, code flexibility, and accuracy. SAM is still in early stages of development and will require continued support from NEAMS to fulfill its potential and to mature into a production tool for advanced reactor safety analysis. The effort to couple SAS4A/SASSYS-1 and SAM is the first step on the integration of these modeling capabilities.« less

Assessment of radiological protection systems among diagnostic radiology facilities in North East India.

PubMed

Singh, Thokchom Dewan; Jayaraman, T; Arunkumar Sharma, B

2017-03-01

This study aims to assess the adequacy level of radiological protection systems available in the diagnostic radiology facilities located in three capital cities of North East (NE) India. It further attempts to understand, using a multi-disciplinary approach, how the safety codes/standards in diagnostic radiology framed by the Atomic Energy Regulatory Board (AERB) and the International Atomic Energy Agency (IAEA) to achieve adequate radiological protection in facilities, have been perceived, conceptualized, and applied accordingly in these facilities. About 30 diagnostic radiology facilities were randomly selected from three capitals of states in NE India; namely Imphal (Manipur), Shillong (Meghalaya) and Guwahati (Assam). A semi-structured questionnaire developed based on a multi-disciplinary approach was used for this study. It was observed that radiological practices undertaken in these facilities were not exactly in line with safety codes/standards in diagnostic radiology of the AERB and the IAEA. About 50% of the facilities had registered/licensed x-ray equipment with the AERB. More than 80% of the workers did not use radiation protective devices, although these devices were available in the facilities. About 85% of facilities had no institutional risk management system. About 70% of the facilities did not carry out periodic quality assurance testing of their x-ray equipment or surveys of radiation leakage around the x-ray room, and did not display radiation safety indicators in the x-ray rooms. Workers in these facilities exhibited low risk perception about the risks associated with these practices. The majority of diagnostic radiology facilities in NE India did not comply with the radiological safety codes/standards framed by the AERB and IAEA. The study found inadequate levels of radiological protection systems in the majority of facilities. This study suggests a need to establish firm measures that comply with the radiological safety codes/standards of the AERB and IAEA to protect patients, workers and the public of this region.

Applications of a damage tolerance analysis methodology in aircraft design and production

NASA Technical Reports Server (NTRS)

Woodward, M. R.; Owens, S. D.; Law, G. E.; Mignery, L. A.

1992-01-01

Objectives of customer mandated aircraft structural integrity initiatives in design are to guide material selection, to incorporate fracture resistant concepts in the design, to utilize damage tolerance based allowables and planned inspection procedures necessary to enhance the safety and reliability of manned flight vehicles. However, validated fracture analysis tools for composite structures are needed to accomplish these objectives in a timely and economical manner. This paper briefly describes the development, validation, and application of a damage tolerance methodology for composite airframe structures. A closed-form analysis code, entitled SUBLAM was developed to predict the critical biaxial strain state necessary to cause sublaminate buckling-induced delamination extension in an impact damaged composite laminate. An embedded elliptical delamination separating a thin sublaminate from a thick parent laminate is modelled. Predicted failure strains were correlated against a variety of experimental data that included results from compression after impact coupon and element tests. An integrated analysis package was developed to predict damage tolerance based margin-of-safety (MS) using NASTRAN generated loads and element information. Damage tolerance aspects of new concepts are quickly and cost-effectively determined without the need for excessive testing.

Hazard Analysis for Pneumatic Flipper Suitport/Z-1 Manned Evaluation, Chamber B, Building 32. Revision: Basic

NASA Technical Reports Server (NTRS)

2012-01-01

One of the characteristics of an effective safety program is the recognition and control of hazards before mishaps or failures occur. Conducting potentially hazardous tests necessitates a thorough hazard analysis in order to protect our personnel from injury and our equipment from damage. The purpose of this hazard analysis is to define and address the potential hazards and controls associated with the Z1 Suit Port Test in Chamber B located in building 32, and to provide the applicable team of personnel with the documented results. It is imperative that each member of the team be familiar with the hazards and controls associated with his/her particular tasks, assignments, and activities while interfacing with facility test systems, equipment, and hardware. The goal of this hazard analysis is to identify all hazards that have the potential to harm personnel and/or damage facility equipment, flight hardware, property, or harm the environment. This analysis may also assess the significance and risk, when applicable, of lost test objectives when substantial monetary value is involved. The hazards, causes, controls, verifications, and risk assessment codes have been documented on the hazard analysis work sheets in appendix A of this document. The preparation and development of this report is in accordance with JPR 1700.1, JSC Safety and Health Handbook.

78 FR 2939 - Adoption of Recommendations

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-15

...) (excluding penalties under the Internal Revenue Code, the Occupational Safety and Health Act, the Social... research reveals that the statute is an undesirably blunt tool for reducing the duplicative burdens that... risk to public health, safety, or other important values, third-party program rules should guarantee a...

30 CFR 56.12047 - Guy wires.

Code of Federal Regulations, 2010 CFR

2010-07-01

... insulator protection of the National Electrical Safety Code, part 2, entitled “Safety Rules for the Installation and Maintenance of Electric Supply and Communication Lines” (also referred to as National Bureau... from the National Institute of Science and Technology, 100 Bureau Drive, Stop 3460, Gaithersburg, MD...

30 CFR 57.12047 - Guy wires.

Code of Federal Regulations, 2010 CFR

2010-07-01

... requirements for grounding or insulator protection of the National Electrical Safety Code, part 2, entitled “Safety Rules for the Installation and Maintenance of Electric Supply and Communication Lines” (also... documents may be obtained from the National Institute of Science and Technology, 100 Bureau Drive, Stop 3460...

30 CFR 57.12047 - Guy wires.

Code of Federal Regulations, 2012 CFR

2012-07-01

... requirements for grounding or insulator protection of the National Electrical Safety Code, part 2, entitled “Safety Rules for the Installation and Maintenance of Electric Supply and Communication Lines” (also... documents may be obtained from the National Institute of Science and Technology, 100 Bureau Drive, Stop 3460...

30 CFR 57.12047 - Guy wires.

Code of Federal Regulations, 2013 CFR

2013-07-01

... requirements for grounding or insulator protection of the National Electrical Safety Code, part 2, entitled “Safety Rules for the Installation and Maintenance of Electric Supply and Communication Lines” (also... documents may be obtained from the National Institute of Science and Technology, 100 Bureau Drive, Stop 3460...

30 CFR 57.12047 - Guy wires.

Code of Federal Regulations, 2014 CFR

2014-07-01

... requirements for grounding or insulator protection of the National Electrical Safety Code, part 2, entitled “Safety Rules for the Installation and Maintenance of Electric Supply and Communication Lines” (also... documents may be obtained from the National Institute of Science and Technology, 100 Bureau Drive, Stop 3460...

29 CFR 1952.117 - Changes to approved plans.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR... the amendment on October 24, 1988. (2) The State submitted amendments to its Occupational Safety and Health Act (chapter 69, Utah Code Annotated 1953), which became effective on April 29, 1985, which...

Eigenvalue Contributon Estimator for Sensitivity Calculations with TSUNAMI-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T; Williams, Mark L

2007-01-01

Since the release of the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) codes in SCALE [1], the use of sensitivity and uncertainty analysis techniques for criticality safety applications has greatly increased within the user community. In general, sensitivity and uncertainty analysis is transitioning from a technique used only by specialists to a practical tool in routine use. With the desire to use the tool more routinely comes the need to improve the solution methodology to reduce the input and computational burden on the user. This paper reviews the current solution methodology of the Monte Carlo eigenvalue sensitivity analysismore » sequence TSUNAMI-3D, describes an alternative approach, and presents results from both methodologies.« less

Comprehensive target populations for current active safety systems using national crash databases.

PubMed

Kusano, Kristofer D; Gabler, Hampton C

2014-01-01

The objective of active safety systems is to prevent or mitigate collisions. A critical component in the design of active safety systems is the identification of the target population for a proposed system. The target population for an active safety system is that set of crashes that a proposed system could prevent or mitigate. Target crashes have scenarios in which the sensors and algorithms would likely activate. For example, the rear-end crash scenario, where the front of one vehicle contacts another vehicle traveling in the same direction and in the same lane as the striking vehicle, is one scenario for which forward collision warning (FCW) would be most effective in mitigating or preventing. This article presents a novel set of precrash scenarios based on coded variables from NHTSA's nationally representative crash databases in the United States. Using 4 databases (National Automotive Sampling System-General Estimates System [NASS-GES], NASS Crashworthiness Data System [NASS-CDS], Fatality Analysis Reporting System [FARS], and National Motor Vehicle Crash Causation Survey [NMVCCS]) the scenarios developed in this study can be used to quantify the number of police-reported crashes, seriously injured occupants, and fatalities that are applicable to proposed active safety systems. In this article, we use the precrash scenarios to identify the target populations for FCW, pedestrian crash avoidance systems (PCAS), lane departure warning (LDW), and vehicle-to-vehicle (V2V) or vehicle-to-infrastructure (V2I) systems. Crash scenarios were derived using precrash variables (critical event, accident type, precrash movement) present in all 4 data sources. This study found that these active safety systems could potentially mitigate approximately 1 in 5 of all severity and serious injury crashes in the United States and 26 percent of fatal crashes. Annually, this corresponds to 1.2 million all severity, 14,353 serious injury (MAIS 3+), and 7412 fatal crashes. In addition, we provide the source code for the crash scenarios as an appendix (see online supplement) to this article so that researchers can use the crash scenarios in future research.

Increased Rail Transit Vehicle Crashworthiness in Head-On Collisions. Volume IV. TRAIN User's Manual.

DOT National Transportation Integrated Search

1980-06-01

A specific goal of safety is to reduce the number of injuries that may result from the collision of two trains. In Volume IV, a computer code for the simulated crash of two railcar consists is described. The code is capable of simulating the mechanic...

National Electrical Code in Power Engineering Course for Electrical Engineering Curriculum

ERIC Educational Resources Information Center

Azizur, Rahman M. M.

2011-01-01

In order to ensure the safety of their inhabitants and properties, the residential, industrial and business installations require complying with NEC (national electrical code) for electrical systems. Electrical design engineers and technicians rely heavily on these very important design guidelines. However, these design guidelines are not formally…

7 CFR 1755.503 - General.

Code of Federal Regulations, 2010 CFR

2010-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2012 CFR

2012-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2011 CFR

2011-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2013 CFR

2013-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

7 CFR 1755.503 - General.

Code of Federal Regulations, 2014 CFR

2014-01-01

.../Institute of Electrical and Electronics Engineers, Inc. (ANSI/IEEE) C2-1997, National Electrical Safety Code (NESC). The National Electrical Code ® and NEC ® are registered trademarks of the National Fire... have been made by the RUS borrower or the engineer delegated by the RUS borrower. (f) Only a qualified...

Fire Protection in Educational Occupancies.

ERIC Educational Resources Information Center

Gervais, Romeo P.

2000-01-01

Discusses the origins of school fires and the components of the fire protection code called the Life Safety Code (LSC). Three of the following LSC requirements are described: means of egress; protection from hazards; and fire suppression and alarm systems. Information on who starts fires is highlighted along with preventive measures. (GR)

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 4 2013-10-01 2013-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 4 2014-10-01 2014-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

Keeping a common bawdy house becomes a "serious offence" under Criminal Code.

PubMed

2010-10-01

New federal regulations targeting organized crime will make keeping a common bawdy house a "serious offence" under the Criminal Code. Sex work advocates reacted by calling the measure a serious step back that will undermine the protection of sex workers' human rights, safety, dignity and health.

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doucet, M.; Durant Terrasson, L.; Mouton, J.

2006-07-01

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recentlymore » updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)« less

Generating Code Review Documentation for Auto-Generated Mission-Critical Software

NASA Technical Reports Server (NTRS)

Denney, Ewen; Fischer, Bernd

2009-01-01

Model-based design and automated code generation are increasingly used at NASA to produce actual flight code, particularly in the Guidance, Navigation, and Control domain. However, since code generators are typically not qualified, there is no guarantee that their output is correct, and consequently auto-generated code still needs to be fully tested and certified. We have thus developed AUTOCERT, a generator-independent plug-in that supports the certification of auto-generated code. AUTOCERT takes a set of mission safety requirements, and formally verifies that the autogenerated code satisfies these requirements. It generates a natural language report that explains why and how the code complies with the specified requirements. The report is hyper-linked to both the program and the verification conditions and thus provides a high-level structured argument containing tracing information for use in code reviews.

DUKSUP: A Computer Program for High Thrust Launch Vehicle Trajectory Design and Optimization

NASA Technical Reports Server (NTRS)

Williams, C. H.; Spurlock, O. F.

2014-01-01

From the late 1960's through 1997, the leadership of NASA's Intermediate and Large class unmanned expendable launch vehicle projects resided at the NASA Lewis (now Glenn) Research Center (LeRC). One of LeRC's primary responsibilities --- trajectory design and performance analysis --- was accomplished by an internally-developed analytic three dimensional computer program called DUKSUP. Because of its Calculus of Variations-based optimization routine, this code was generally more capable of finding optimal solutions than its contemporaries. A derivation of optimal control using the Calculus of Variations is summarized including transversality, intermediate, and final conditions. The two point boundary value problem is explained. A brief summary of the code's operation is provided, including iteration via the Newton-Raphson scheme and integration of variational and motion equations via a 4th order Runge-Kutta scheme. Main subroutines are discussed. The history of the LeRC trajectory design efforts in the early 1960's is explained within the context of supporting the Centaur upper stage program. How the code was constructed based on the operation of the Atlas/Centaur launch vehicle, the limits of the computers of that era, the limits of the computer programming languages, and the missions it supported are discussed. The vehicles DUKSUP supported (Atlas/Centaur, Titan/Centaur, and Shuttle/Centaur) are briefly described. The types of missions, including Earth orbital and interplanetary, are described. The roles of flight constraints and their impact on launch operations are detailed (such as jettisoning hardware on heating, Range Safety, ground station tracking, and elliptical parking orbits). The computer main frames on which the code was hosted are described. The applications of the code are detailed, including independent check of contractor analysis, benchmarking, leading edge analysis, and vehicle performance improvement assessments. Several of DUKSUP's many major impacts on launches are discussed including Intelsat, Voyager, Pioneer Venus, HEAO, Galileo, and Cassini.

DUKSUP: A Computer Program for High Thrust Launch Vehicle Trajectory Design and Optimization

NASA Technical Reports Server (NTRS)

Spurlock, O. Frank; Williams, Craig H.

2015-01-01

From the late 1960s through 1997, the leadership of NASAs Intermediate and Large class unmanned expendable launch vehicle projects resided at the NASA Lewis (now Glenn) Research Center (LeRC). One of LeRCs primary responsibilities --- trajectory design and performance analysis --- was accomplished by an internally-developed analytic three dimensional computer program called DUKSUP. Because of its Calculus of Variations-based optimization routine, this code was generally more capable of finding optimal solutions than its contemporaries. A derivation of optimal control using the Calculus of Variations is summarized including transversality, intermediate, and final conditions. The two point boundary value problem is explained. A brief summary of the codes operation is provided, including iteration via the Newton-Raphson scheme and integration of variational and motion equations via a 4th order Runge-Kutta scheme. Main subroutines are discussed. The history of the LeRC trajectory design efforts in the early 1960s is explained within the context of supporting the Centaur upper stage program. How the code was constructed based on the operation of the AtlasCentaur launch vehicle, the limits of the computers of that era, the limits of the computer programming languages, and the missions it supported are discussed. The vehicles DUKSUP supported (AtlasCentaur, TitanCentaur, and ShuttleCentaur) are briefly described. The types of missions, including Earth orbital and interplanetary, are described. The roles of flight constraints and their impact on launch operations are detailed (such as jettisoning hardware on heating, Range Safety, ground station tracking, and elliptical parking orbits). The computer main frames on which the code was hosted are described. The applications of the code are detailed, including independent check of contractor analysis, benchmarking, leading edge analysis, and vehicle performance improvement assessments. Several of DUKSUPs many major impacts on launches are discussed including Intelsat, Voyager, Pioneer Venus, HEAO, Galileo, and Cassini.

Validation and verification of the laser range safety tool (LRST)

NASA Astrophysics Data System (ADS)

Kennedy, Paul K.; Keppler, Kenneth S.; Thomas, Robert J.; Polhamus, Garrett D.; Smith, Peter A.; Trevino, Javier O.; Seaman, Daniel V.; Gallaway, Robert A.; Crockett, Gregg A.

2003-06-01

The U.S. Dept. of Defense (DOD) is currently developing and testing a number of High Energy Laser (HEL) weapons systems. DOD range safety officers now face the challenge of designing safe methods of testing HEL's on DOD ranges. In particular, safety officers need to ensure that diffuse and specular reflections from HEL system targets, as well as direct beam paths, are contained within DOD boundaries. If both the laser source and the target are moving, as they are for the Airborne Laser (ABL), a complex series of calculations is required and manual calculations are impractical. Over the past 5 years, the Optical Radiation Branch of the Air Force Research Laboratory (AFRL/HEDO), the ABL System Program Office, Logicon-RDA, and Northrup-Grumman, have worked together to develop a computer model called teh Laser Range Safety Tool (LRST), specifically designed for HEL reflection hazard analyses. The code, which is still under development, is currently tailored to support the ABL program. AFRL/HEDO has led an LRST Validation and Verification (V&V) effort since 1998, in order to determine if code predictions are accurate. This paper summarizes LRST V&V efforts to date including: i) comparison of code results with laboratory measurements of reflected laser energy and with reflection measurements made during actual HEL field tests, and ii) validation of LRST's hazard zone computations.