10 CFR 52.137 - Contents of applications; technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

... limits on its operation, and presents a safety analysis of the structures, systems, and components and of... products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation...

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.

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

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

G. S. Chang

2007-09-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.« less

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.

Main steam-line break core shroud loading calculations for BWRs

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

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1995-12-31

In July 1994, the U.S. Nuclear regulatory Commission sent out Generic Letter 94-03 to all boiling water reactors in the United States, informing them of intergranular stress corrosion cracking of core shrouds found in 2 reactors. The letter directed all to perform safety analysis of the BWR units. Penn State performed scoping calculations to determine the forces experienced by the core shroud during a main-stream line break transient.

The Influence of Health Policy and Market Factors on the Hospital Safety Net

PubMed Central

Bazzoli, Gloria J; Lindrooth, Richard C; Kang, R ay; Hasnain-Wynia, R omana

2006-01-01

Objective To examine how the financial pressures resulting from the Balanced Budget Act (BBA) of 1997 interacted with private sector pressures to affect indigent care provision. Data Sources/Study Setting American Hospital Association Annual Survey, Area Resource File, InterStudy Health Maintenance Organization files, Current Population Survey, and Bureau of Primary Health Care data. Study Design We distinguished core and voluntary safety net hospitals in our analysis. Core safety net hospitals provide a large share of uncompensated care in their markets and have large indigent care patient mix. Voluntary safety net hospitals provide substantial indigent care but less so than core hospitals. We examined the effect of financial pressure in the initial year of the 1997 BBA on uncompensated care for three hospital groups. Data for 1996–2000 were analyzed using approaches that control for hospital and market heterogeneity. Data Collection/Extraction Methods All urban U.S. general acute care hospitals with complete data for at least 2 years between 1996 and 2000, which totaled 1,693 institutions. Principal Findings Core safety net hospitals reduced their uncompensated care in response to Medicaid financial pressure. Voluntary safety net hospitals also responded in this way but only when faced with the combined forces of Medicaid and private sector payment pressures. Nonsafety net hospitals did not exhibit similar responses. Conclusions Our results are consistent with theories of hospital behavior when institutions face reductions in payment. They raise concern given continuing state budget crises plus the focus of recent federal deficit reduction legislation intended to cut Medicaid expenditures. PMID:16899001

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Identification of Core Competencies for an Undergraduate Food Safety Curriculum Using a Modified Delphi Approach

ERIC Educational Resources Information Center

Johnston, Lynette M.; Wiedmann, Martin; Orta-Ramirez, Alicia; Oliver, Haley F.; Nightingale, Kendra K.; Moore, Christina M.; Stevenson, Clinton D.; Jaykus, Lee-Ann

2014-01-01

Identification of core competencies for undergraduates in food safety is critical to assure courses and curricula are appropriate in maintaining a well-qualified food safety workforce. The purpose of this study was to identify and refine core competencies relevant to postsecondary food safety education using a modified Delphi method. Twenty-nine…

Gap analysis: a method to assess core competency development in the curriculum.

PubMed

Fater, Kerry H

2013-01-01

To determine the extent to which safety and quality improvement core competency development occurs in an undergraduate nursing program. Rapid change and increased complexity of health care environments demands that health care professionals are adequately prepared to provide high quality, safe care. A gap analysis compared the present state of competency development to a desirable (ideal) state. The core competencies, Nurse of the Future Nursing Core Competencies, reflect the ideal state and represent minimal expectations for entry into practice from pre-licensure programs. Findings from the gap analysis suggest significant strengths in numerous competency domains, deficiencies in two competency domains, and areas of redundancy in the curriculum. Gap analysis provides valuable data to direct curriculum revision. Opportunities for competency development were identified, and strategies were created jointly with the practice partner, thereby enhancing relevant knowledge, attitudes, and skills nurses need for clinical practice currently and in the future.

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.

Biomechanical Evaluation of a Tooth Restored with High Performance Polymer PEKK Post-Core System: A 3D Finite Element Analysis.

PubMed

Lee, Ki-Sun; Shin, Joo-Hee; Kim, Jong-Eun; Kim, Jee-Hwan; Lee, Won-Chang; Shin, Sang-Wan; Lee, Jeong-Yol

2017-01-01

The aim of this study was to evaluate the biomechanical behavior and long-term safety of high performance polymer PEKK as an intraradicular dental post-core material through comparative finite element analysis (FEA) with other conventional post-core materials. A 3D FEA model of a maxillary central incisor was constructed. A cyclic loading force of 50 N was applied at an angle of 45° to the longitudinal axis of the tooth at the palatal surface of the crown. For comparison with traditionally used post-core materials, three materials (gold, fiberglass, and PEKK) were simulated to determine their post-core properties. PEKK, with a lower elastic modulus than root dentin, showed comparably high failure resistance and a more favorable stress distribution than conventional post-core material. However, the PEKK post-core system showed a higher probability of debonding and crown failure under long-term cyclic loading than the metal or fiberglass post-core systems.

Biomechanical Evaluation of a Tooth Restored with High Performance Polymer PEKK Post-Core System: A 3D Finite Element Analysis

PubMed Central

Shin, Joo-Hee; Kim, Jong-Eun; Kim, Jee-Hwan; Lee, Won-Chang; Shin, Sang-Wan

2017-01-01

The aim of this study was to evaluate the biomechanical behavior and long-term safety of high performance polymer PEKK as an intraradicular dental post-core material through comparative finite element analysis (FEA) with other conventional post-core materials. A 3D FEA model of a maxillary central incisor was constructed. A cyclic loading force of 50 N was applied at an angle of 45° to the longitudinal axis of the tooth at the palatal surface of the crown. For comparison with traditionally used post-core materials, three materials (gold, fiberglass, and PEKK) were simulated to determine their post-core properties. PEKK, with a lower elastic modulus than root dentin, showed comparably high failure resistance and a more favorable stress distribution than conventional post-core material. However, the PEKK post-core system showed a higher probability of debonding and crown failure under long-term cyclic loading than the metal or fiberglass post-core systems. PMID:28386547

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

NASA Astrophysics Data System (ADS)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

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.

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

ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

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

Chang, G.S.

2008-07-15

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores. (author)« less

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

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

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

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

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

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

Brunett, A. J.; Fei, T.; Strons, P. S.

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort ismore » to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.« less

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Tank 241-AP-105, cores 208, 209 and 210, analytical results for the final report

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

Nuzum, J.L.

1997-10-24

This document is the final laboratory report for Tank 241-AP-105. Push mode core segments were removed from Risers 24 and 28 between July 2, 1997, and July 14, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-AP-105 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997) and Tank Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT), differential scanning calorimetry (DSC) analysis, or total organic carbon (TOC) analysis exceeded the notification limits as stated in TSAP and DQO. The statisticalmore » results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report. Appearance and Sample Handling Two cores, each consisting of four segments, were expected from Tank 241-AP-105. Three cores were sampled, and complete cores were not obtained. TSAP states core samples should be transported to the laboratory within three calendar days from the time each segment is removed from the tank. This requirement was not met for all cores. Attachment 1 illustrates subsamples generated in the laboratory for analysis and identifies their sources. This reference also relates tank farm identification numbers to their corresponding 222-S Laboratory sample numbers.« less

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

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

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

Nuclear reactor safety research since three mile island.

PubMed

Mynatt, F R

1982-04-09

The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

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

Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh

2015-12-21

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spotsmore » in the VHTR core.« less

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

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

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Tank 241-T-204, core 188 analytical results for the final report

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

Nuzum, J.L.

TANK 241-T-204, CORE 188, ANALYTICAL RESULTS FOR THE FINAL REPORT. This document is the final laboratory report for Tank 241 -T-204. Push mode core segments were removed from Riser 3 between March 27, 1997, and April 11, 1997. Segments were received and extruded at 222-8 Laboratory. Analyses were performed in accordance with Tank 241-T-204 Push Mode Core Sampling and analysis Plan (TRAP) (Winkleman, 1997), Letter of instruction for Core Sample Analysis of Tanks 241-T-201, 241- T-202, 241-T-203, and 241-T-204 (LAY) (Bell, 1997), and Safety Screening Data Qual@ Objective (DO) ODukelow, et al., 1995). None of the subsamples submitted for totalmore » alpha activity (AT) or differential scanning calorimetry (DC) analyses exceeded the notification limits stated in DO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group and are not considered in this report.« less

Tenofovir Containing Thiolated Chitosan Core/Shell Nanofibers: In Vitro and in Vivo Evaluations.

PubMed

Meng, Jianing; Agrahari, Vivek; Ezoulin, Miezan J; Zhang, Chi; Purohit, Sudhaunshu S; Molteni, Agostino; Dim, Daniel; Oyler, Nathan A; Youan, Bi-Botti C

2016-12-05

It is hypothesized that thiolated chitosan (TCS) core/shell nanofibers (NFs) can enhance the drug loading of tenofovir, a model low molecular weight and highly water-soluble drug molecule, and improve its mucoadhesivity and in vivo safety. To test this hypothesis, poly(ethylene oxide) (PEO) core with TCS and polylactic acid (PLA) shell NFs are fabricated by a coaxial electrospinning technique. The morphology, drug loading, drug release profiles, cytotoxicity and mucoadhesion of the NFs are analyzed using scanning and transmission electron microscopies, liquid chromatography, cytotoxicity assays on VK2/E6E7 and End1/E6E7 cell lines and Lactobacilli crispatus, fluorescence imaging and periodic acid colorimetric method, respectively. In vivo safety studies are performed in C57BL/6 mice followed by H&E and immunohistochemical (CD45) staining analysis of genital tract. The mean diameters of PEO, PEO/TCS, and PEO/TCS-PLA NFs are 118.56, 9.95, and 99.53 nm, respectively. The NFs exhibit smooth surface. The drug loading (13%-25%, w/w) increased by 10-fold compared to a nanoparticle formulation due to the application of the electrospinning technique. The NFs are noncytotoxic at the concentration of 1 mg/mL. The PEO/TCS-PLA core/shell NFs mostly exhibit a release kinetic following Weibull model (r 2 = 0.9914), indicating the drug release from a matrix system. The core/shell NFs are 40-60-fold more bioadhesive than the pure PEO based NFs. The NFs are nontoxic and noninflammatory in vivo after daily treatment for up to 7 days. Owing to their enhanced drug loading and preliminary safety profile, the TCS core/shell NFs are promising candidates for the topical delivery of HIV/AIDS microbicides such as tenofovir.

Hydrogen Safety Project: Chemical analysis support task. Window ``E`` analyses

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

Jones, T E; Campbell, J A; Hoppe, E W

1992-09-01

Core samples taken from tank 101-SY at Hanford during ``window E`` were analyzed for organic and radiochemical constituents by staff of the Analytical Chemistry Laboratory at Pacific Northwest Laboratory. Westinghouse Hanford company submitted these samples to the laboratory.

Approach to numerical safety guidelines based on a core melt criterion. [PWR; BWR

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

Azarm, M.A.; Hall, R.E.

1982-01-01

A plausible approach is proposed for translating a single level criterion to a set of numerical guidelines. The criterion for core melt probability is used to set numerical guidelines for various core melt sequences, systems and component unavailabilities. These guidelines can be used as a means for making decisions regarding the necessity for replacing a component or improving part of a safety system. This approach is applied to estimate a set of numerical guidelines for various sequences of core melts that are analyzed in Reactor Safety Study for the Peach Bottom Nuclear Power Plant.

The Accident at TEPCO's Fukushima-Daiichi Nuclear Power Plant: Technical Description of What Happened and Lessons Learned for the Future

NASA Astrophysics Data System (ADS)

Omoto, Akira

2012-02-01

Tsunami that followed M9.0 earthquake on March 11^th left the Fukushima-Daiichi Nuclear Power Plants without power and heat sink. While water makeup continued by AC-independent systems to keep the fuel core covered by coolant, operating team tried to depressurize and enable low pressure injection to the reactor to avoid overheating but was not successful enough primarily due to limited available resources. This resulted in core melt, hydrogen explosion and release of radioactivity to the environment. Key lessons learned are; 1) safety regulation and safety culture, 2) workable/executable severe accident management procedure, 3) crisis management and 4) design. Implications on security include revealed vulnerability and the nexus of safety and security. Given the scale of damage to the environmental, attention must be paid to defense against it and to societal safety goal of nuclear power by considering offsite remedial costs, compensation to damage, energy replacement cost etc. A sort of root cause analysis first by asking ``Why nuclear community failed to prevent this accident?'' was initiated by the University of Tokyo.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

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

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

Applying Qualitative Hazard Analysis to Support Quantitative Safety Analysis for Proposed Reduced Wake Separation Conops

NASA Technical Reports Server (NTRS)

Shortle, John F.; Allocco, Michael

2005-01-01

This paper describes a scenario-driven hazard analysis process to identify, eliminate, and control safety-related risks. Within this process, we develop selective criteria to determine the applicability of applying engineering modeling to hypothesized hazard scenarios. This provides a basis for evaluating and prioritizing the scenarios as candidates for further quantitative analysis. We have applied this methodology to proposed concepts of operations for reduced wake separation for closely spaced parallel runways. For arrivals, the process identified 43 core hazard scenarios. Of these, we classified 12 as appropriate for further quantitative modeling, 24 that should be mitigated through controls, recommendations, and / or procedures (that is, scenarios not appropriate for quantitative modeling), and 7 that have the lowest priority for further analysis.

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

Impulse Flashover Tests at Edgar Beauchamp High Voltage Test Facility, Dixon, California, in Support of Cutler Insulator Failure Investigation

DTIC Science & Technology

2006-07-01

sites. The strength member of the safety core insulators is a fiberglass belt wrapped around pins in the end fittings. Porcelain tubes cover the belt... porcelain tube and heavily tracked the fiberglass belt but left the belt intact structurally (Figure 1). Figure 1. Cutler safety core insulator ...fail-safe insulators . For these tests, the porcelain tube of the safety core insulator was replaced with a plastic see-through tube. The test report [5

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

Public Safety Core. Integrated Academic and Technical Competencies (ITAC).

ERIC Educational Resources Information Center

Ohio State Dept. of Education, Columbus. Div. of Career-Technical and Adult Education.

This document, which lists the public safety core competencies that are part of the Integrated Academic and Technical Competencies (ITAC) in Ohio, is intended to assist individuals and organizations develop a course to provide students with knowledge and skills applicable to public safety careers, including but not limited to firefighter,…

Implementation of a patient safety program at a tertiary health system: A longitudinal analysis of interventions and serious safety events.

PubMed

Cropper, Douglas P; Harb, Nidal H; Said, Patricia A; Lemke, Jon H; Shammas, Nicolas W

2018-04-01

We hypothesize that implementation of a safety program based on high reliability organization principles will reduce serious safety events (SSE). The safety program focused on 7 essential elements: (a) safety rounding, (b) safety oversight teams, (c) safety huddles, (d) safety coaches, (e) good catches/safety heroes, (f) safety education, and (g) red rule. An educational curriculum was implemented focusing on changing high-risk behaviors and implementing critical safety policies. All unusual occurrences were captured in the Midas system and investigated by risk specialists, the safety officer, and the chief medical officer. A multidepartmental committee evaluated these events, and a root cause analysis (RCA) was performed. Events were tabulated and serious safety event (SSE) recorded and plotted over time. Safety success stories (SSSs) were also evaluated over time. A steady drop in SSEs was seen over 9 years. Also a rise in SSSs was evident, reflecting on staff engagement in the program. The parallel change in SSEs, SSSs, and the implementation of various safety interventions highly suggest that the program was successful in achieving its goals. A safety program based on high-reliability organization principles and made a core value of the institution can have a significant positive impact on reducing SSEs. © 2018 American Society for Healthcare Risk Management of the American Hospital Association.

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

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Automatic safety rod for reactors

DOEpatents

Germer, John H.

1988-01-01

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

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

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

The incidence of coring with blunt versus sharp needles.

PubMed

Wani, Tariq; Wadhwa, Anupama; Tobias, Joseph D

2014-03-01

With the advent of safety needles to prevent inadvertent needle sticks in the operating room (OR), a potentially new issue has arisen. These needles may result in coring, or the shaving off of fragments of the rubber stopper, when the needle is pierced through the rubber stopper of the medication vial. These fragments may be left in the vial and then drawn up with the medication and possibly injected into patients. The current study prospectively evaluated the incidence of coring when blunt and sharp needles were used to pierce rubber topped vials. We also evaluated the incidence of coring in empty medication vials with rubber tops. The rubber caps were then pierced with either an18-gauge sharp hypodermic needle or a blunt plastic (safety) needle. Coring occurred in 102 of 250 (40.8%) vials when a blunt needle was used versus 9 of 215 (4.2%) vials with a sharp needle (P < 0.0001). A significant incidence of coring was demonstrated when a blunt plastic safety needle was used. This situation is potentially a patient safety hazard and methods to eliminate this problem are needed. Copyright © 2014 Elsevier Inc. All rights reserved.

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.

Scale Model Test and Transient Analysis of Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Ohmori, Shuichi; Narabayashi, Tadashi; Mori, Michitsugu

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feed-water heater that heats up feed-water by using extracted steam from turbine. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying "high-efficiency SI", which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feed-water heaters and emergency core cooling system of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a severe accident-free concept). This paper describes the results of the scale model test, and the transient analysis of SI-driven passive core injection system (PCIS).

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

Semiquantitative analysis of gaps in microbiological performance of fish processing sector implementing current food safety management systems: a case study.

PubMed

Onjong, Hillary Adawo; Wangoh, John; Njage, Patrick Murigu Kamau

2014-08-01

Fish processing plants still face microbial food safety-related product rejections and the associated economic losses, although they implement legislation, with well-established quality assurance guidelines and standards. We assessed the microbial performance of core control and assurance activities of fish exporting processors to offer suggestions for improvement using a case study. A microbiological assessment scheme was used to systematically analyze microbial counts in six selected critical sampling locations (CSLs). Nine small-, medium- and large-sized companies implementing current food safety management systems (FSMS) were studied. Samples were collected three times on each occasion (n = 324). Microbial indicators representing food safety, plant and personnel hygiene, and overall microbiological performance were analyzed. Microbiological distribution and safety profile levels for the CSLs were calculated. Performance of core control and assurance activities of the FSMS was also diagnosed using an FSMS diagnostic instrument. Final fish products from 67% of the companies were within the legally accepted microbiological limits. Salmonella was absent in all CSLs. Hands or gloves of workers from the majority of companies were highly contaminated with Staphylococcus aureus at levels above the recommended limits. Large-sized companies performed better in Enterobacteriaceae, Escherichia coli, and S. aureus than medium- and small-sized ones in a majority of the CSLs, including receipt of raw fish material, heading and gutting, and the condition of the fish processing tables and facilities before cleaning and sanitation. Fish products of 33% (3 of 9) of the companies and handling surfaces of 22% (2 of 9) of the companies showed high variability in Enterobacteriaceae counts. High variability in total viable counts and Enterobacteriaceae was noted on fish products and handling surfaces. Specific recommendations were made in core control and assurance activities associated with sampling locations showing poor performance.

Quality and Safety as a Core Leadership Competency.

PubMed

Bleich, Michael R

2018-05-01

A leader's toolbox of competencies comprises knowledge, skills, and abilities in clinical care, finance, human resource management, and more. As essential as these are, a strong command of quality and safety competencies is sovereign in leading and managing, ensuring an optimal patient experience. Four core areas of quality and safety competencies are presented: systems science, knowledge workers, implementation science and big data, and quality safety tools and techniques. J Contin Educ Nurs. 2018;49(5):200-202. Copyright 2018, SLACK Incorporated.

Educating the ambulance technician, paramedic, and clinical supervisor: using factor analysis to inform the curriculum

PubMed Central

Kilner, T

2004-01-01

Methods: Data generated by a Delphi study investigating the desirable attributes of ambulance technician, paramedic, and clinical supervisor were subject to factor analysis to explore inter-relations between the variables or desirable attributes. Variables that loaded onto any factor at a correlation level of >0.3 were included in the analysis. Results: Three factors emerged in each of the occupational groups. In respect of the ambulance technician these factors may be described as; core professional skills, individual and collaborative approaches to health and safety, and the management of self and clinical situations. For the paramedic the themes are; core professional skills, management of self and clinical situations, and approaches to health and safety. For the clinical supervisor there is again a theme described as core professional skills, with a further two themes described as role model and lifelong learning. Conclusions: The profile of desirable attributes emerging from this study are remarkably similar to the generic benchmark statements for health care programmes outlined by the Quality Assurance Agency for Higher Education. It seems that a case is emerging for a revision of the curriculum currently used for the education and training of ambulance staff, which is more suited to a consumer led health service and which reflects the broader professional base seen in programmes associated with other healthcare professions. This study has suggested outline content, and module structure for the education of the technician, paramedic, and clinical supervisor, based on empirical evidence. PMID:15107389

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« 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

Comparisons of Wilks’ and Monte Carlo Methods in Response to the 10CFR50.46(c) Proposed Rulemaking

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

Zhang, Hongbin; Szilard, Ronaldo; Zou, Ling

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10CFR50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently nearly all the fuel rods in a reactor core need to be analyzed under LOCA conditions to demonstrate compliance to the safety limits. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of themore » limiting rods as well as to quantify the safety margins under LOCA conditions. With the new analysis method presented in this work, the limiting transient case and the limiting rods can be easily identified to quantify the safety margins in response to the proposed new rulemaking. In this work, the best-estimate plus uncertainty (BEPU) analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. Both the direct Monte Carlo method and the Wilks’ nonparametric statistical method can be used to perform uncertainty quantification. Wilks’ method has become the de-facto industry standard to perform uncertainty quantification in BEPU LOCA analyses. Despite its widespread adoption by the industry, the use of small sample sizes to infer statement of compliance to the existing 10CFR50.46 rule, has been a major cause of unrealized operational margin in today’s BEPU methods. Moreover the debate on the proper interpretation of the Wilks’ theorem in the context of safety analyses is not fully resolved yet, even more than two decades after its introduction in the frame of safety analyses in the nuclear industry. This represents both a regulatory and application risk in rolling out new methods. With the 10CFR50.46(c) proposed rulemaking, the deficiencies of the Wilks’ approach are further exacerbated. The direct Monte Carlo approach offers a robust alternative to perform uncertainty quantification within the context of BEPU analyses. In this work, the Monte Carlo method is compared with the Wilks’ method in response to the NRC 10CFR50.46(c) proposed rulemaking.« less

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

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

Garner, P. L.; Hanan, N. A.

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decidemore » to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.« less

"What--me worry?" "Why so serious?": a personal view on the Fukushima nuclear reactor accidents.

PubMed

Gallucci, Raymond

2012-09-01

Infrequently, it seems that a significant accident precursor or, worse, an actual accident, involving a commercial nuclear power reactor occurs to remind us of the need to reexamine the safety of this important electrical power technology from a risk perspective. Twenty-five years since the major core damage accident at Chernobyl in the Ukraine, the Fukushima reactor complex in Japan experienced multiple core damages as a result of an earthquake-induced tsunami beyond either the earthquake or tsunami design basis for the site. Although the tsunami itself killed tens of thousands of people and left the area devastated and virtually uninhabitable, much concern still arose from the potential radioactive releases from the damaged reactors, even though there was little population left in the area to be affected. As a lifelong probabilistic safety analyst in nuclear engineering, even I must admit to a recurrence of the doubt regarding nuclear power safety after Fukushima that I had experienced after Three Mile Island and Chernobyl. This article is my attempt to "recover" my personal perspective on acceptable risk by examining both the domestic and worldwide history of commercial nuclear power plant accidents and attempting to quantify the risk in terms of the frequency of core damage that one might glean from a review of operational history. © 2012 Society for Risk Analysis.

NASA Tech Briefs, June 2011

NASA Technical Reports Server (NTRS)

2011-01-01

Topics covered include: Wind and Temperature Spectrometry of the Upper Atmosphere in Low-Earth Orbit; Health Monitor for Multitasking, Safety-Critical, Real-Time Software; Stereo Imaging Miniature Endoscope; Early Oscillation Detection Technique for Hybrid DC/DC Converters; Parallel Wavefront Analysis for a 4D Interferometer; Schottky Heterodyne Receivers With Full Waveguide Bandwidth; Carbon Nanofiber-Based, High-Frequency, High-Q, Miniaturized Mechanical Resonators; Ultracapacitor-Based Uninterrupted Power Supply System; Coaxial Cables for Martian Extreme Temperature Environments; Using Spare Logic Resources To Create Dynamic Test Points; Autonomous Coordination of Science Observations Using Multiple Spacecraft; Autonomous Phase Retrieval Calibration; EOS MLS Level 1B Data Processing Software, Version 3; Cassini Tour Atlas Automated Generation; Software Development Standard Processes (SDSP); Graphite Composite Panel Polishing Fixture; Material Gradients in Oxygen System Components Improve Safety; Ridge Waveguide Structures in Magnesium-Doped Lithium Niobate; Modifying Matrix Materials to Increase Wetting and Adhesion; Lightweight Magnetic Cooler With a Reversible Circulator; The Invasive Species Forecasting System; Method for Cleanly and Precisely Breaking Off a Rock Core Using a Radial Compressive Force; Praying Mantis Bending Core Breakoff and Retention Mechanism; Scoring Dawg Core Breakoff and Retention Mechanism; Rolling-Tooth Core Breakoff and Retention Mechanism; Vibration Isolation and Stabilization System for Spacecraft Exercise Treadmill Devices; Microgravity-Enhanced Stem Cell Selection; Diagnosis and Treatment of Neurological Disorders by Millimeter-Wave Stimulation; Passive Vaporizing Heat Sink; Remote Sensing and Quantization of Analog Sensors; Phase Retrieval for Radio Telescope and Antenna Control; Helium-Cooled Black Shroud for Subscale Cryogenic Testing; Receive Mode Analysis and Design of Microstrip Reflectarrays; and Chance-Constrained Guidance With Non-Convex Constraints.

Maintenance Resources by Building Use for U.S. Army Installations. Volume 3. Appendices I through P

DTIC Science & Technology

1991-05-01

floor I____ ______ __ Solid core (safety 0423230 10 2.78 glass ) painted exterior door Drinking fountain 081 1HOO _______6 1.77 Wood, finished 0415FI0...Concrete, finished 062B200 25 1.15 flooring Solid core wood 0423230 26 1.12 (safety glass ) paint exterior door Safety switch, 1122100 27 1.07 enclosed...exterior, 1st floor Steel frame 0431210 26 1.04 (painted) operable window, 1st floor Steel (w/safety 0421220 27 1.04 glass ) painted exterior door Pipe

Fight Bac! | Partnership for Food Safety Education

Science.gov Websites

Games & Activities School Lunches Free Resources Be Food Safe Resources Brand Assets Brochures & Spanish Free Resources For Consumers, Retailers and Educators The Four Core Practices Food Safety Basics workers and parents Free Resources The Core Four Food Poisoning Child Care Training schoolchildren Hands

Identification of core functions and development of a deployment planning tool for safety service patrols in Virginia.

DOT National Transportation Integrated Search

2006-01-01

The purpose of this study was to identify and document the core functions of the Virginia Department of Transportation's (VDOT) Safety Service Patrol (SSP) programs and to develop a deployment planning tool that would help VDOT decision-makers when c...

Chicago Monostatic Acoustic Vortex Sensing System. Volume IV. Wake Vortex Decay.

DTIC Science & Technology

1982-07-01

analysis here, the peak velocity core radius cannot be directly compared to the present results. If one applies the analysis of Table 10 to the LDV vortex...Tietjens, O.G., Applied Hydro- and Aeromechanics, Dover, New York, 1957, pp. 158-163. 11. Hallock, J.N., "Vortex Advisory System Safety Analysis, Vol. I...Stability and Control Characteristics Model DC-9-30 Jet Transport," LB-32323, Dec. 1966 (revised Oct. 1968), Douglas Aircraft Company , Long Beach, CA. 13

Automatic safety rod for reactors. [LMFBR

DOEpatents

Germer, J.H.

1982-03-23

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

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

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

Ranking of sabotage/tampering avoidance technology alternatives

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

Andrews, W.B.; Tabatabai, A.S.; Powers, T.B.

1986-01-01

Pacific Northwest Laboratory conducted a study to evaluate alternatives to the design and operation of nuclear power plants, emphasizing a reduction of their vulnerability to sabotage. Estimates of core melt accident frequency during normal operations and from sabotage/tampering events were used to rank the alternatives. Core melt frequency for normal operations was estimated using sensitivity analysis of results of probabilistic risk assessments. Core melt frequency for sabotage/tampering was estimated by developing a model based on probabilistic risk analyses, historic data, engineering judgment, and safeguards analyses of plant locations where core melt events could be initiated. Results indicate the most effectivemore » alternatives focus on large areas of the plant, increase safety system redundancy, and reduce reliance on single locations for mitigation of transients. Less effective options focus on specific areas of the plant, reduce reliance on some plant areas for safe shutdown, and focus on less vulnerable targets.« less

Self-Sustaining Thorium Boiling Water Reactors

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

Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare themore » RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.« less

Nine-Year Core Study Data for Sientra's FDA-Approved Round and Shaped Implants with High-Strength Cohesive Silicone Gel.

PubMed

Stevens, W Grant; Calobrace, M Bradley; Harrington, Jennifer; Alizadeh, Kaveh; Zeidler, Kamakshi R; d'Incelli, Rosalyn C

2016-04-01

Since approval in March 2012, data on Sientra's (Santa Barbara, CA) silicone gel implants have been updated and published regularly to provide immediate visibility to the continued safety and performance of these devices. The 9 year follow-up data support the previously published data confirming the ongoing safety and efficacy of Sientra silicone gel breast implants. The authors provide updated 9 year study data for Sientra's round and shaped silicone gel breast implants. The Core Study is an ongoing 10 year study that enrolled 1788 patients with 3506 Sientra implants across four indications (primary augmentation, revision-augmentation, primary reconstruction, and revision-reconstruction). For the safety analysis, Kaplan-Meier risk rates were calculated to evaluate postoperative complications, including all breast implant-related adverse effects. For the effectiveness analyses, results were presented through 8 years as patient satisfaction scores were assessed at even years. Through 9 years, the overall risk of capsular contracture was 12.6%. Smooth devices (16.6%, 95% CI, 14.2%, 19.5%) had a statistically significantly higher rate of capsular contracture compared to textured devices (8.0%, 95% CI, 6.2%, 10.4%). Out of the 610 reoperations in 477 patients, over half of all reoperations were due to cosmetic reasons (n = 315; 51.6%). Patient satisfaction remains high through 8 years, with 90% of primary augmentation patients indicating their breast implants look natural and feel soft. The 9-year follow-up data from the ongoing Core Study of the Sientra portfolio of HSC and HSC+ silicone gel breast implants reaffirm the very strong safety profile as well as continued patient satisfaction. 2 Therapeutic. © 2016 The American Society for Aesthetic Plastic Surgery, Inc. Reprints and permission: journals.permissions@oup.com.

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

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

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Diagnostic performance and safety of a three-dimensional 14-core systematic biopsy method.

PubMed

Takeshita, Hideki; Kawakami, Satoru; Numao, Noboru; Sakura, Mizuaki; Tatokoro, Manabu; Yamamoto, Shinya; Kijima, Toshiki; Komai, Yoshinobu; Saito, Kazutaka; Koga, Fumitaka; Fujii, Yasuhisa; Fukui, Iwao; Kihara, Kazunori

2015-03-01

To investigate the diagnostic performance and safety of a three-dimensional 14-core biopsy (3D14PBx) method, which is a combination of the transrectal six-core and transperineal eight-core biopsy methods. Between December 2005 and August 2010, 1103 men underwent 3D14PBx at our institutions and were analysed prospectively. Biopsy criteria included a PSA level of 2.5-20 ng/mL or abnormal digital rectal examination (DRE) findings, or both. The primary endpoint of the study was diagnostic performance and the secondary endpoint was safety. We applied recursive partitioning to the entire study cohort to delineate the unique contribution of each sampling site to overall and clinically significant cancer detection. Prostate cancer was detected in 503 of the 1103 patients (45.6%). Age, family history of prostate cancer, DRE, PSA, percentage of free PSA and prostate volume were associated with the positive biopsy results significantly and independently. Of the 503 cancers detected, 39 (7.8%) were clinically locally advanced (≥cT3a), 348 (69%) had a biopsy Gleason score (GS) of ≥7, and 463 (92%) met the definition of biopsy-based significant cancer. Recursive partitioning analysis showed that each sampling site contributed uniquely to both the overall and the biopsy-based significant cancer detection rate of the 3D14PBx method. The overall cancer-positive rate of each sampling site ranged from 14.5% in the transrectal far lateral base to 22.8% in the transrectal far lateral apex. As of August 2010, 210 patients (42%) had undergone radical prostatectomy, of whom 55 (26%) were found to have pathologically non-organ-confined disease, 174 (83%) had prostatectomy GS ≥7 and 185 (88%) met the definition of prostatectomy-based significant cancer. This is the first prospective analysis of the diagnostic performance of an extended biopsy method, which is a simplified version of the somewhat redundant super-extended three-dimensional 26-core biopsy. As expected, each sampling site uniquely contributed not only to overall cancer detection, but also to significant cancer detection. 3D14PBx is a feasible systematic biopsy method in men with PSA <20 ng/mL. © 2014 The Authors. BJU International © 2014 BJU International.

Ending on a positive: Examining the role of safety leadership decisions, behaviours and actions in a safety critical situation.

PubMed

Donovan, Sarah-Louise; Salmon, Paul M; Horberry, Timothy; Lenné, Michael G

2018-01-01

Safety leadership is an important factor in supporting safe performance in the workplace. The present case study examined the role of safety leadership during the Bingham Canyon Mine high-wall failure, a significant mining incident in which no fatalities or injuries were incurred. The Critical Decision Method (CDM) was used in conjunction with a self-reporting approach to examine safety leadership in terms of decisions, behaviours and actions that contributed to the incidents' safe outcome. Mapping the analysis onto Rasmussen's Risk Management Framework (Rasmussen, 1997), the findings demonstrate clear links between safety leadership decisions, and emergent behaviours and actions across the work system. Communication and engagement based decisions featured most prominently, and were linked to different leadership practices across the work system. Further, a core sub-set of CDM decision elements were linked to the open flow and exchange of information across the work system, which was critical to supporting the safe outcome. The findings provide practical implications for the development of safety leadership capability to support safety within the mining industry. Copyright © 2017 Elsevier Ltd. All rights reserved.

Basic Safety II. Apprentice Related Training Module.

ERIC Educational Resources Information Center

Rice, Eric; Spetz, Sally H.

One in a series of core instructional materials for apprentices to use during the first or second years of apprentice-related subjects training, this booklet deals with basic safety. The first section consists of an outline of the content and scope of the core materials as well as a self-assessment pretest. Covered in the four instructional…

Nursing physical assessment for patient safety in general wards: reaching consensus on core skills.

PubMed

Douglas, Clint; Booker, Catriona; Fox, Robyn; Windsor, Carol; Osborne, Sonya; Gardner, Glenn

2016-07-01

To determine consensus across acute care specialty areas on core physical assessment skills necessary for early recognition of changes in patient status in general wards. Current approaches to physical assessment are inconsistent and have not evolved to meet increased patient and system demands. New models of nursing assessment are needed in general wards that ensure a proactive and patient safety approach. A modified Delphi study. Focus group interviews with 150 acute care registered nurses at a large tertiary referral hospital generated a framework of core skills that were developed into a web-based survey. We then sought consensus with a panel of 35 senior acute care registered nurses following a classical Delphi approach over three rounds. Consensus was predefined as at least 80% agreement for each skill across specialty areas. Content analysis of focus group transcripts identified 40 discrete core physical assessment skills. In the Delphi rounds, 16 of these were consensus validated as core skills and were conceptually aligned with the primary survey: (Airway) Assess airway patency; (Breathing) Measure respiratory rate, Evaluate work of breathing, Measure oxygen saturation; (Circulation) Palpate pulse rate and rhythm, Measure blood pressure by auscultation, Assess urine output; (Disability) Assess level of consciousness, Evaluate speech, Assess for pain; (Exposure) Measure body temperature, Inspect skin integrity, Inspect and palpate skin for signs of pressure injury, Observe any wounds, dressings, drains and invasive lines, Observe ability to transfer and mobilise, Assess bowel movements. Among a large and diverse group of experienced acute care registered nurses consensus was achieved on a structured core physical assessment to detect early changes in patient status. Although further research is needed to refine the model, clinical application should promote systematic assessment and clinical reasoning at the bedside. © 2016 John Wiley & Sons Ltd.

Safety and Security Interface Technology Initiative

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

Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

Safety and Security Interface Technology Initiative Mr. Kevin J. Carroll Dr. Robert Lowrie, Dr. Micheal Lehto BWXT Y12 NSC Oak Ridge, TN 37831 865-576-2289/865-241-2772 carrollkj@y12.doe.gov Work Objective. Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. “Supporting Excellence in Operations Through Safety Analysis,” (workshop theme)more » includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is “Safeguards/Security Integration with Safety.” This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis/Security Documentation Integration, Configuration Control, and development of a shared ‘tool box’ of information/successes. Specific Benefits. The expectation or end state resulting from the topical report and associated implementation plan includes: (1) A recommended process for handling the documentation of the security and safety disciplines, including an appropriate change control process and participation by all stakeholders. (2) A means to package security systems with sufficient information to help expedite the flow of that system through the process. In addition, a means to share successes among sites, to include information and safety basis to the extent such information is transportable. (3) Identification of key security systems and associated essential security elements being installed and an arrangement for the sites installing these systems to host an appropriate team to review a specific system and determine what information is exportable. (4) Identification of the security systems’ essential elements and appropriate controls required for testing of these essential elements in the facility. (5) The ability to help refine and improve an agreed to control set at the manufacture stage.« less

Diagnostic accuracy of 22/25-gauge core needle in endoscopic ultrasound-guided sampling: systematic review and meta-analysis.

PubMed

Oh, Hyoung-Chul; Kang, Hyun; Lee, Jae Young; Choi, Geun Joo; Choi, Jung Sik

2016-11-01

To compare the diagnostic accuracy of endoscopic ultrasound-guided core needle aspiration with that of standard fine-needle aspiration by systematic review and meta-analysis. Studies using 22/25-gauge core needles, irrespective of comparison with standard fine needles, were comprehensively reviewed. Pooled sensitivity, specificity, diagnostic odds ratio (DOR), and summary receiver operating characteristic curves for the diagnosis of malignancy were used to estimate the overall diagnostic efficiency. The pooled sensitivity, specificity, and DOR of the core needle for the diagnosis of malignancy were 0.88 (95% confidence interval [CI], 0.84 to 0.90), 0.99 (95% CI, 0.96 to 1), and 167.37 (95% CI, 65.77 to 425.91), respectively. The pooled sensitivity, specificity, and DOR of the standard needle were 0.84 (95% CI, 0.79 to 0.88), 1 (95% CI, 0.97 to 1), and 130.14 (95% CI, 34.00 to 495.35), respectively. The area under the curve of core and standard needle in the diagnosis of malignancy was 0.974 and 0.955, respectively. The core and standard needle were comparable in terms of pancreatic malignancy diagnosis. There was no significant difference in procurement of optimal histologic cores between core and standard needles (risk ratio [RR], 0.545; 95% CI, 0.187 to 1.589). The number of needle passes for diagnosis was significantly lower with the core needle (standardized mean difference, -0.72; 95% CI, -1.02 to -0.41). There were no significant differences in overall complications (RR, 1.26; 95% CI, 0.34 to 4.62) and technical failure (RR, 5.07; 95% CI, 0.68 to 37.64). Core and standard needles were comparable in terms of diagnostic accuracy, technical performance, and safety profile.

Spaceborne power systems preference analyses. Volume 2: Decision analysis

NASA Technical Reports Server (NTRS)

Smith, J. H.; Feinberg, A.; Miles, R. F., Jr.

1985-01-01

Sixteen alternative spaceborne nuclear power system concepts were ranked using multiattribute decision analysis. The purpose of the ranking was to identify promising concepts for further technology development and the issues associated with such development. Four groups were interviewed to obtain preference. The four groups were: safety, systems definition and design, technology assessment, and mission analysis. The highest ranked systems were the heat-pipe thermoelectric systems, heat-pipe Stirling, in-core thermionic, and liquid-metal thermoelectric systems. The next group contained the liquid-metal Stirling, heat-pipe Alkali Metal Thermoelectric Converter (AMTEC), heat-pipe Brayton, liquid-metal out-of-core thermionic, and heat-pipe Rankine systems. The least preferred systems were the liquid-metal AMTEC, heat-pipe thermophotovoltaic, liquid-metal Brayton and Rankine, and gas-cooled Brayton. The three nonheat-pipe technologies selected matched the top three nonheat-pipe systems ranked by this study.

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

Campbell, J.A.; Clauss, S.A.; Grant, K.E.

The objectives of this task are to develop and document extraction and analysis methods for organics in waste tanks, and to extend these methods to the analysis of actual core samples to support the Waste Tank organic Safety Program. This report documents progress at Pacific Northwest Laboratory (a) during FY 1994 on methods development, the analysis of waste from Tank 241-C-103 (Tank C-103) and T-111, and the transfer of documented, developed analytical methods to personnel in the Analytical Chemistry Laboratory (ACL) and 222-S laboratory. This report is intended as an annual report, not a completed work.

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

Culture matters: indigenizing patient safety in Bhutan.

PubMed

Pelzang, Rinchen; Johnstone, Megan-Jane; Hutchinson, Alison M

2017-09-01

Studies show that if quality of healthcare in a country is to be achieved, due consideration must be given to the importance of the core cultural values as a critical factor in improving patient safety outcomes. The influence of Bhutan's traditional (core) cultural values on the attitudes and behaviours of healthcare professionals regarding patient care are not known. This study aimed to explore the possible influence of Bhutan's traditional cultural values on staff attitudes towards patient safety and quality care. Undertaken as a qualitative exploratory descriptive inquiry, a purposeful sample of 94 healthcare professionals and managers were recruited from three levels of hospitals, a training institute and the Ministry of Health. Interviews were transcribed verbatim and analysed using thematic analysis strategies. The findings of the study suggest that Bhutanese traditional cultural values have both productive and counterproductive influences on staff attitudes towards healthcare delivery and the processes that need to be in place to ensure patient safety. Productive influences encompassed: karmic incentives to avoid preventable harm and promote safe patient care; and the prospective adoption of the 'four harmonious friends' as a culturally meaningful frame for improving understanding of the role and importance of teamwork in enhancing patient safety. Counterproductive influences included: the adoption of hierarchical and authoritative styles of management; unilateral decision-making; the legitimization of karmic beliefs; differential treatment of patients; and preferences for traditional healing practices and rituals. Although problematic in some areas, Bhutan's traditional cultural values could be used positively to inform and frame an effective model for improving patient safety in Bhutan's hospitals. Such a model must entail the institution of an 'indigenized' patient safety program, with patient safety research and reporting systems framed around local patient safety concerns and solutions, including religious and cultural concepts, values and perspectives. © The Author 2017. Published by Oxford University Press in association with The London School of Hygiene and Tropical Medicine. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

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

Bacchiani, M.; Medich, C.; Rigamonti, M.

1995-09-01

The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2more » test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.« less

Detection of QT prolongation using a novel ECG analysis algorithm applying intelligent automation: Prospective blinded evaluation using the Cardiac Safety Research Consortium ECG database

PubMed Central

Green, Cynthia L.; Kligfield, Paul; George, Samuel; Gussak, Ihor; Vajdic, Branislav; Sager, Philip; Krucoff, Mitchell W.

2013-01-01

Background The Cardiac Safety Research Consortium (CSRC) provides both “learning” and blinded “testing” digital ECG datasets from thorough QT (TQT) studies annotated for submission to the US Food and Drug Administration (FDA) to developers of ECG analysis technologies. This manuscript reports the first results from a blinded “testing” dataset that examines Developer re-analysis of original Sponsor-reported core laboratory data. Methods 11,925 anonymized ECGs including both moxifloxacin and placebo arms of a parallel-group TQT in 191 subjects were blindly analyzed using a novel ECG analysis algorithm applying intelligent automation. Developer measured ECG intervals were submitted to CSRC for unblinding, temporal reconstruction of the TQT exposures, and statistical comparison to core laboratory findings previously submitted to FDA by the pharmaceutical sponsor. Primary comparisons included baseline-adjusted interval measurements, baseline- and placebo-adjusted moxifloxacin QTcF changes (ddQTcF), and associated variability measures. Results Developer and Sponsor-reported baseline-adjusted data were similar with average differences less than 1 millisecond (ms) for all intervals. Both Developer and Sponsor-reported data demonstrated assay sensitivity with similar ddQTcF changes. Average within-subject standard deviation for triplicate QTcF measurements was significantly lower for Developer than Sponsor-reported data (5.4 ms and 7.2 ms, respectively; p<0.001). Conclusion The virtually automated ECG algorithm used for this analysis produced similar yet less variable TQT results compared to the Sponsor-reported study, without the use of a manual core laboratory. These findings indicate CSRC ECG datasets can be useful for evaluating novel methods and algorithms for determining QT/QTc prolongation by drugs. While the results should not constitute endorsement of specific algorithms by either CSRC or FDA, the value of a public domain digital ECG warehouse to provide prospective, blinded comparisons of ECG technologies applied for QT/QTc measurement is illustrated. PMID:22424006

Common Core Curriculum for Vocational Education. Category F: Stages and Structure of Curriculum Development. F-4: Safety.

ERIC Educational Resources Information Center

Winegar, Gary

This module on safety is one of a set of four on stages and structure of curriculum development and is part of a larger series of thirty-four modules comprising a core curriculum intended for use in the professional preparation of vocational educators in the areas of agricultural, business, home economics, and industrial education. Following the…

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

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

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

Estimation of Ultimate Tensile Strength of dentin Using Finite Element Analysis from Endodontically Treated Tooth

NASA Astrophysics Data System (ADS)

Sinthaworn, S.; Puengpaiboon, U.; Warasetrattana, N.; Wanapaisarn, S.

2018-01-01

Endodontically treated teeth were simulated by finite element analysis in order to estimate ultimate tensile strength of dentin. Structures of the endodontically treated tooth cases are flared root canal, restored with different number of fiber posts {i.e. resin composite core without fiber post (group 1), fiber post No.3 with resin composite core (group 2) and fiber post No.3 accessory 2 fiber posts No.0 with resin composite core (group 3)}. Elastic modulus and Poisson’s ratio of materials were selected from literatures. The models were loaded by the average fracture resistances load of each groups (group 1: 361.80 N, group 2: 559.46 N, group 3: 468.48 N) at 135 degree angulation in respect to the longitudinal axis of the teeth. The stress analysis and experimental confirm that fracture zone is at dentin area. To estimate ultimate tensile strength of dentin, trial and error of ultimate tensile strength were tested to obtain factor of safety (FOS) equal to 1.00. The result reveals that ultimate tensile strength of dentin of group 1, 2, 3 are 38.89, 30.96, 37.19 MPa, respectively.

Analysis of the return to power scenario following a LBLOCA in a PWR

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

Macian, R.; Tyler, T.N.; Mahaffy, J.H.

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus,more » the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.« less

Patient safety education at Japanese medical schools: results of a nationwide survey.

PubMed

Maeda, Shoichi; Kamishiraki, Etsuko; Starkey, Jay

2012-05-10

Patient safety education, including error prevention strategies and management of adverse events, has become a topic of worldwide concern. The importance of the patient safety is also recognized in Japan following two serious medical accidents in 1999. Furthermore, educational curriculum guideline revisions in 2008 by relevant the Ministry of Education includes patient safety as part of the core medical curriculum. However, little is known about the patient safety education in Japanese medical schools partly because a comprehensive study has not yet been conducted in this field. Therefore, we have conducted a nationwide survey in order to clarify the current status of patient safety education at medical schools in Japan. Response rate was 60.0% (n = 48/80). Ninety-eight-percent of respondents (n = 47/48) reported integration of patient safety education into their curricula. Thirty-nine percent reported devoting less than five hours to the topic. All schools that teach patient safety reported use of lecture based teaching methods while few used alternative methods, such as role-playing or in-hospital training. Topics related to medical error theory and legal ramifications of error are widely taught while practical topics related to error analysis such as root cause analysis are less often covered. Based on responses to our survey, most Japanese medical schools have incorporated the topic of patient safety into their curricula. However, the number of hours devoted to the patient safety education is far from the sufficient level with forty percent of medical schools that devote five hours or less to it. In addition, most medical schools employ only the lecture based learning, lacking diversity in teaching methods. Although most medical schools cover basic error theory, error analysis is taught at fewer schools. We still need to make improvements to our medical safety curricula. We believe that this study has the implications for the rest of the world as a model of what is possible and a sounding board for what topics might be important.

76 FR 44081 - Agency Information Collection Activities: Notice of Request for Approval of a New Information...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-22

...: Jeffrey Miller, (202) 366-0744 or [email protected] , Office of Safety Integration, Federal Highway... Friday, except Federal holidays. SUPPLEMENTARY INFORMATION: Title: Strategic Highway Safety Plan (SHSP... Highway Safety Improvement Program (HSIP) as a core Federal program. A Strategic Highway Safety Plan (SHSP...

Directory of Academic Programs in Occupational Safety and Health.

ERIC Educational Resources Information Center

Weis, William J., III; And Others

This booklet describes academic program offerings in American colleges and universities in the area of occupational safety and health. Programs are divided into five major categories, corresponding to each of the core disciplines: (1) occupational safety and health/industrial hygiene, (2) occupational safety, (3) industrial hygiene, (4)…

Safety. Unit 8: A Core Curriculum of Related Instruction for Apprentices.

ERIC Educational Resources Information Center

New York State Education Dept., Albany. Bureau of Occupational and Career Curriculum Development.

The safety education unit is presented to assist apprentices to acquire a general knowledge of procedures for insuring safety on the job. The unit consists of 10 modules: (1) the Occupational Safety and Health Act: safety and health bill of rights for workers; (2) accident prevention; (3) first aid; (4) accident reports; importance, use, and how…

Primer on the Highway Safety Improvement Program (HSIP)

DOT National Transportation Integrated Search

2014-09-16

The Highway Safety Improvement Program (HSIP) is a core Federal-aid program for State Departments of Transportation (State DOTs) administered by the Federal Highway Administration (FHWA). This is a major source of funding for safety projects on the n...

Street Intersection Characteristics and Their Impacts on Perceived Bicycling Safety

DOT National Transportation Integrated Search

2018-01-01

Safety concern is one of the core issues that deter people from bicycling in the US. Earlier studies have explored the associations between intersection design characteristics and bicyclist safety perceptions. Research shows that there are significan...

Determination of the core temperature of a Li-ion cell during thermal runaway

NASA Astrophysics Data System (ADS)

Parhizi, M.; Ahmed, M. B.; Jain, A.

2017-12-01

Safety and performance of Li-ion cells is severely affected by thermal runaway where exothermic processes within the cell cause uncontrolled temperature rise, eventually leading to catastrophic failure. Most past experimental papers on thermal runaway only report surface temperature measurement, while the core temperature of the cell remains largely unknown. This paper presents an experimentally validated method based on thermal conduction analysis to determine the core temperature of a Li-ion cell during thermal runaway using surface temperature and chemical kinetics data. Experiments conducted on a thermal test cell show that core temperature computed using this method is in good agreement with independent thermocouple-based measurements in a wide range of experimental conditions. The validated method is used to predict core temperature as a function of time for several previously reported thermal runaway tests. In each case, the predicted peak core temperature is found to be several hundreds of degrees Celsius higher than the measured surface temperature. This shows that surface temperature alone is not sufficient for thermally characterizing the cell during thermal runaway. Besides providing key insights into the fundamental nature of thermal runaway, the ability to determine the core temperature shown here may lead to practical tools for characterizing and mitigating thermal runaway.

Detection of QT prolongation using a novel electrocardiographic analysis algorithm applying intelligent automation: prospective blinded evaluation using the Cardiac Safety Research Consortium electrocardiographic database.

PubMed

Green, Cynthia L; Kligfield, Paul; George, Samuel; Gussak, Ihor; Vajdic, Branislav; Sager, Philip; Krucoff, Mitchell W

2012-03-01

The Cardiac Safety Research Consortium (CSRC) provides both "learning" and blinded "testing" digital electrocardiographic (ECG) data sets from thorough QT (TQT) studies annotated for submission to the US Food and Drug Administration (FDA) to developers of ECG analysis technologies. This article reports the first results from a blinded testing data set that examines developer reanalysis of original sponsor-reported core laboratory data. A total of 11,925 anonymized ECGs including both moxifloxacin and placebo arms of a parallel-group TQT in 181 subjects were blindly analyzed using a novel ECG analysis algorithm applying intelligent automation. Developer-measured ECG intervals were submitted to CSRC for unblinding, temporal reconstruction of the TQT exposures, and statistical comparison to core laboratory findings previously submitted to FDA by the pharmaceutical sponsor. Primary comparisons included baseline-adjusted interval measurements, baseline- and placebo-adjusted moxifloxacin QTcF changes (ddQTcF), and associated variability measures. Developer and sponsor-reported baseline-adjusted data were similar with average differences <1 ms for all intervals. Both developer- and sponsor-reported data demonstrated assay sensitivity with similar ddQTcF changes. Average within-subject SD for triplicate QTcF measurements was significantly lower for developer- than sponsor-reported data (5.4 and 7.2 ms, respectively; P < .001). The virtually automated ECG algorithm used for this analysis produced similar yet less variable TQT results compared with the sponsor-reported study, without the use of a manual core laboratory. These findings indicate that CSRC ECG data sets can be useful for evaluating novel methods and algorithms for determining drug-induced QT/QTc prolongation. Although the results should not constitute endorsement of specific algorithms by either CSRC or FDA, the value of a public domain digital ECG warehouse to provide prospective, blinded comparisons of ECG technologies applied for QT/QTc measurement is illustrated. Copyright © 2012 Mosby, Inc. All rights reserved.

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

NASA Astrophysics Data System (ADS)

Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

NASA Astrophysics Data System (ADS)

Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

2018-02-01

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

Safe sleep, day and night: mothers' experiences regarding infant sleep safety.

PubMed

Lau, Annie; Hall, Wendy

2016-10-01

To explore Canadian mothers' experiences with infant sleep safety. Parents decide when, how and where to place their infants to sleep. It is anticipated that they will follow international Sudden Infant Death Syndrome prevention sleep safety guidelines. Limited evidence is available for how parents take up guidelines; no studies have explored Canadian mothers' experiences regarding infant sleep safety. An inductive qualitative descriptive study using some elements of grounded theory, including concurrent data collection and analysis and memoing. Semi-structured interviews and constant comparative analysis were employed to explore infant sleep safety experiences of 14 Canadian mothers residing in Metro Vancouver. Data collection commenced in December 2012 and ended in July 2013. The core theme, Infant Sleep Safety Cycle, represents a cyclical process encompassing sleep safety from the prenatal period to the first six months of infants' lives. The cyclical process includes five segments: mothers' expectations of sleep safety, their struggles with reality as opposed to maternal visions, modifications of expectations, provision of rationale for choices and shifts in mothers' views of infants' developmental capabilities. Mothers' experiences were influenced by four factors: perceptions of everyone's needs, familial influences, attitudes and judgments from outsiders and resource availability and accessibility. To manage infants' sleep, mothers reframed sleep safety guidelines and downplayed the risk of Sudden Infant Death Syndrome for all forms of sleep at all times. Healthcare providers can support mothers' efforts to manage their infants' sleep challenges. During prenatal and postpartum periods, providers' interventions can influence mothers' efforts to adhere to sleep safety principles. The study findings support healthcare providers' efforts to assist mothers to modify expectations and develop strategies to support sleep safety principles while acknowledging their challenges. © 2016 John Wiley & Sons Ltd.

The relationship between patient safety climate and standard precaution adherence: a systematic review of the literature

PubMed Central

Hessels, Amanda; Larson, Elaine

2015-01-01

SUMMARY Standard precaution (SP) adherence is universally suboptimal, despite being a core component of healthcare-associated infection (HCAI) prevention and healthcare worker (HCW) safety. Emerging evidence suggests that patient safety climate (PSC) factors may improve HCW behaviours. Our aim was to examine the relationship between PSC and SP adherence by HCWs in acute care hospitals. A systematic review was conducted as guided by the Preferred Reporting Items for Systematic Reviews and Meta-Analysis. Three electronic databases were comprehensively searched for literature published or available in English between 2000 and 2014. Seven of 888 articles identified were eligible for final inclusion in the review. Two reviewers independently assessed study quality using a validated quality tool. The seven articles were assigned quality scores ranging from 7 to 10 of 10 possible points. Five measured all aspects of SP and two solely measured needlestick and sharps handling. Three included a secondary outcome of HCW exposure; none included HCAIs. All reported a statistically significant relationship between better PSC and greater SP adherence and used data from self-report surveys including validated PSC measures or measures of management support and leadership. Although limited in number, studies were of high quality and confirmed that PSC and SP adherence were correlated, suggesting that efforts to improve PSC may enhance adherence to a core component of HCAI prevention and HCW safety. More clearly evident is the need for additional high-quality research. PMID:26549480

Safety monitoring and reactor transient interpreter

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

Hench, J. E.; Fukushima, T. Y.

1983-12-20

An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18,20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24)more » for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18,20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.« less

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

Brown, C. S.; Zhang, Hongbin

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Industry Initiated Core Safety Attributes for Human Spaceflight for the 7th IAASS Conference

NASA Technical Reports Server (NTRS)

Mango, Edward J.

2014-01-01

Now that the NASA Commercial Crew Program (CCP) is beginning its full certification contract for crew transportation to the International Space Station (ISS), is it time for industry to embrace a minimum set of core safety attributes? Those attributes can then be evolved into an industry-led set of basic safety standards and requirements. After 50 years of human space travel sponsored by governments, there are two basic conditions that now exist within the international space industry. The first, there is enough of a space-faring history to encourage the space industry to design, develop and operate human spaceflight systems without government contracts for anything other than services. Second, industry is capable of defining and enforcing a set of industry-based safety attributes and standards for human spaceflight to low-Earth orbit (LEO). This paper will explore both of these basic conditions with a focus on the safety attributes and standards. In the United States, the Federal Aviation Administration (FAA) is now starting to dialogue with industry about the basic safety principles and attributes needed for potential future regulatory oversight. This process is not yet formalized and will take a number of years once approval is given to move forward. Therefore, throughout the next few years, it is an excellent time and opportunity for industry to collaborate together and develop the core set of attributes and standards. As industry engages and embraces a common set of safety attributes, then government agencies, like the FAA and NASA can use that industry-based product to strengthen their efforts on a safe commercial spaceflight foundation for the future. As the commercial space industry takes the lead role in establishing core safety attributes, and then enforcing those attributes, the entire planet can move away from governmental control of design and development and let industry expand safe and successful space operations in LEO. At that point the governmental agencies can focus on oversight of the industries' defined standards and enforcement for common welfare of the space-faring populous and overall public safety.

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.

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

NASA Astrophysics Data System (ADS)

Scarlat, Raluca Olga

This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling approach to the PB-FHR annular pebble bed core cooled by fluoride salt mixtures generated a model that is called Pod. Pod. was used to show the resilience of the PB-FHR core to generation of hot spots or cold spots, due to the effect of buoyancy on the flow and temperature distribution in the packed bed. Pod. was used to investigate the PB-FHR response to ATWS transients. Based on the functional requirements for the core, Pod. was used to generate an optimized design of the flow distribution in the core. An analysis of natural circulation loops cooled by single-phase Boussinesq fluids is presented here, in the context of reactor design that relies on natural circulation decay heat removal, and design of scaled experiments. The scaling arguments are established for a transient natural circulation loop, for loops that have long fluid residence time, and negligible contribution of fluid inertia to the momentum equation. The design of integral effects tests for the loss of forced circulation (LOFC) for PB-FHR is discussed. The special case of natural circulation decay heat removal from a pebble bed reactor was analyzed. A way to define the Reynolds number in a multi-dimensional pebble bed was identified. The scaling methodology for replicating pebble bed friction losses using an electrically resistance heated annular pipe and a needle valve was developed. The thermophysical properties of liquid fluoride salts lead to design of systems with low flow velocities, and hence long fluid residence times. A comparison among liquid coolants for the performance of steady state natural circulation heat removal from a pebble bed was performed. Transient natural circulation experimental data with simulant fluids for fluoride salts is given here. The low flow velocity and the relatively high viscosity of the fluoride salts lead to low Reynolds number flows, and a low Reynolds number in conjunction with a sufficiently high coefficient of thermal expansion makes the system susceptible to local buoyancy effects Experiments indicate that slow exchange of stagnant fluid in static legs can play a significant role in the transient response of natural circulation loops. The effect of non-linear temperature profiles on the hot or cold legs or other segments of the flow loop, which may develop during transient scenarios, should be considered when modeling the performance of natural circulation loops. The data provided here can be used for validation of the application of thermal-hydraulic systems codes to the modeling of heat removal by natural circulation with liquid fluoride salts and its simulant fluids.

Findings from the ISMP Medication Safety Self-Assessment for hospitals.

PubMed

Smetzer, Judy L; Vaida, Allen J; Cohen, Michael R; Tranum, Diane; Pittman, Mary A; Armstrong, Carl W

2003-11-01

Hospital medication practices should be assessed, awareness of the characteristics of a safe medication system heightened, and baseline data to identify national priorities established. A cross-sectional survey of U.S. hospitals (N = 6,180) was conducted in May 2000. The survey instrument contained 194 self-assessment items organized into 20 core characteristics and 10 larger domains. Hospitals were asked to voluntarily submit their confidential assessment data to the Institute for Safe Medication Practices (ISMP) for aggregate analysis. A weighting structure was applied to the individual items and used to calculate core characteristic scores, domain scores, and overall self-assessment scores. These scores were then compared to identify areas most in need of improvement. The 1,435 participating hospitals scored highest in domains related to drug storage and distribution; environmental factors; infusion pumps; and medication labeling, packaging, and nomenclature issues. These hospitals scored lowest in domains related to accessible patient information, communication of medication orders, patient education, and quality processes such as double-check systems and organizational culture. Enormous opportunities exist to improve medication safety, especially in domains related to culture, information management, and communication.

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Foundational workplace safety and health competencies for the emerging workforce☆

PubMed Central

Okun, Andrea H.; Guerin, Rebecca J.; Schulte, Paul A.

2016-01-01

Introduction Young workers (aged 15–24) suffer disproportionately from workplace injuries, with a nonfatal injury rate estimated to be two times higher than among workers age 25 or over. These workers make up approximately 9% of the U.S. workforce and studies have shown that nearly 80% of high school students work at some point during high school. Although young worker injuries are a pressing public health problem, the critical knowledge and skills needed to prepare youth for safe and healthy work are missing from most frameworks used to prepare the emerging U.S. workforce. Methods A framework of foundational workplace safety and health knowledge and skills (the NIOSH 8 Core Competencies)was developed based on the Health Belief Model (HBM). Results The proposed NIOSH Core Competencies utilize the HBM to provide a framework for foundational workplace safety and health knowledge and skills. An examination of how these competencies and the HBM apply to actions that workers take to protect themselves is provided. The social and physical environments that influence these actions are also discussed. Conclusions The NIOSH 8 Core Competencies, grounded in one of the most widely used health behavior theories, fill a critical gap in preparing the emerging U.S. workforce to be cognizant of workplace risks. Practical applications Integration of the NIOSH 8 Core Competencies into school curricula is one way to ensure that every young person has the foundational workplace safety and health knowledge and skills to participate in, and benefit from, safe and healthy work. National Safety Council and Elsevier Ltd. All rights reserved. PMID:27846998

The roles and functions of safety professionals in Taiwan: Comparing the perceptions of safety professionals and safety educators.

PubMed

Wu, Tsung-Chih

2011-10-01

The perspectives of both internal and external members have to be considered when developing safety curricula. This study discusses perceptional differences between safety educators (SEs) and safety professionals (SPs) regarding the function of SPs. The findings will serve as a reference framework for the establishment of core safety competencies and the development of safety curricula for SPs. 248 respondents, including both SEs and SPs, completed self-administered questionnaires, which included the 45-item safety function scale (SFS). Nine factors were extracted from the scale using exploratory factor analysis (EFA), namely inspection and research, regulatory tasks, emergency procedures and settlement of damage, management and financial affairs, culture change, problem identification and analysis, developing and implementing solutions, knowledge management, and training and communications. Descriptive statistical results indicated that SPs and SEs hold differing views on the rank of the frequency of safety functions. MANOVA results indicated that SPs' perceptions of developing and implementing solutions, training and communications, inspection and research, and management and financial affairs were significantly higher than that of SEs. On the other hand, SE's perceptions regarding participation in regulatory tasks were significantly higher than those of SPs. Based on these results, the author suggests that a clear communication channel should be established between universities and industry to reduce the gap between the perceptions of SEs and SPs. The results of the study are statistically and practically significant. In addition to serving as a reference for the development of safety curricula, the results are also conducive to the establishment of SP roles and functions. Ultimately the development of more suitable safety curricula would open up employment competition for students who graduate from safety-related programs. SPs, on the other hand, can correctly recognize their roles and functions so as to realize the safety expectations invested in them by organizations. Copyright © 2011 Elsevier Ltd. All rights reserved.

Acquiring Sediment and Element Compositional Changes Based on a Diffuse Reflectance Spectrophotometry Technology from Cores Offshore Southwestern Taiwan

NASA Astrophysics Data System (ADS)

Pan, H. J.; Chen, M. T.

2014-12-01

Heavy summer monsoon rainfall along with typhoon-induced extreme precipitation cause frequent geological hazards that often threaten the human's safety and property in Taiwan. These geological hazards can be triggered by both natural factors, and/or have become deteriorated by perturbations from more and more human activities ever since few thousand years ago. However, due to the limit of instrumental records for observing long-term environmental changes in Taiwan, few evidence exist for distinguishing the human-induced impacts from natural climate change. Here we report a study on a high quality marine sediment core (MD103264) which were retrieved from the high sedimentation rate area from offshore southwestern Taiwan and present evidence for the long-term climate and possibly human-induced environmental changes since the last glacial. We are using the VIS-NIR Diffuse Reflectance Spectrophotometry (DRS) methods to study the cores. Interpreting the VIS-NIR reflectance spectra through the VARIMAX-rotation, principle component analysis (VPCA) helps conducting rapid and inexpensive measurements for acquiring high-resolution biogenic component, clay, and iron oxide mineral compositional data from the cores. We are also using X-Ray Fluorescence (XRF) analysis, which is also useful in determining the element compositional changes in the core. Our studies aim toward understanding the sediment and element compositional changes that reflect the patterns of changes in precipitation and soil erosion on land since the last glacial to the Holocene, during which the human activities (deforestation, agriculture, and land uses change) may have increased drastically. We will report and interpret the preliminary results of the optical analyses of the core.

Factors impacting perceived safety among staff working on mental health wards.

PubMed

Haines, Alina; Brown, Andrew; McCabe, Rhiannah; Rogerson, Michelle; Whittington, Richard

2017-09-01

Safety at work is a core issue for mental health staff working on in-patient units. At present, there is a limited theoretical base regarding which factors may affect staff perceptions of safety. This study attempted to identify which factors affect perceived staff safety working on in-patient mental health wards. A cross-sectional design was employed across 101 forensic and non-forensic mental health wards, over seven National Health Service trusts nationally. Measures included an online staff survey, Ward Features Checklist and recorded incident data. Data were analysed using categorical principal components analysis and ordinal regression. Perceptions of staff safety were increased by ward brightness, higher number of patient beds, lower staff to patient ratios, less dayroom space and more urban views. The findings from this study do not represent common-sense assumptions. Results are discussed in the context of the literature and may have implications for current initiatives aimed at managing in-patient violence and aggression. None. © The Royal College of Psychiatrists 2017. This is an open access article distributed under the terms of the Creative Commons Attribution (CC BY) license.

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

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

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

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

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

Passive MHD Spectroscopy for Enabling Magnetic Reconstructions on Spherical Tokamak Plasmas at General Fusion Inc

NASA Astrophysics Data System (ADS)

O'Shea, Peter; Laberge, Michel; Mossman, Alex; Reynolds, Meritt

2017-10-01

Magnetic reconstructions on lab based plasma injectors at General Fusion relies heavily on edge magnetic (``Bdot'') probes. On plasma experiments built for field compression (PCS) tests, the number and locations of Bdot probes is limited by mechanical constraints. Additional information about the q profiles near the core in our Spector plasmas is obtained using passive MHD spectroscopy. The coaxial helicity injection (CHI) formation process naturally generates hollow current profiles and reversed shear early in each discharge. Central Ohmic heating naturally peaks the current profiles as our plasmas evolve in time, simultaneously reducing the core safety factor, q(0), and reverse shear. As the central, non-monotonic q-profile crosses rational flux surfaces, we observe transient magnetic reconnection events (MRE's) due to the double tearing mode. Modal analysis allows us to infer the q surfaces involved in each burst. The parametric dependence of the timing of MRE's allows us to estimate the continuous time evolution of the core q profile. Combining core MHD spectroscopy with edge magnetic probe measurements greatly enhances our certainty of the overall q profile.

Role of champions in the implementation of patient safety practice change.

PubMed

Soo, Stephanie; Berta, Whitney; Baker, G Ross

2009-01-01

Practitioners of patient safety practice change agree that champions are central to the success of implementation. The clinical champion role is a concept that has been widely promoted yet empirically underdeveloped in health services literature. Questions remain as to who these champions are, what roles they play in patient safety practice change and what contexts serve to facilitate their efforts. This investigation used a multiple-case study design to critically examine the role of champions in the implementation of rapid response teams (RRTs), an innovative complex patient safety intervention, in two large urban acute care facilities. An analysis of interviews with key individuals involved in the RRT implementation process revealed a typology of the patient safety practice champion that extended beyond clinical personnel to include managerial and executive staff. Champions engaged to a varying extent in a number of core activities, including education, advocacy, relationship building and boundary spanning. Individuals became champions both through informal emergence and a combination of formal appointment and informal emergence. By identifying and elaborating upon specific features of the champion role, this study aims to expand the dialogue about champions for patient safety practice change.

Evaluation of Electronic Healthcare Databases for Post-Marketing Drug Safety Surveillance and Pharmacoepidemiology in China.

PubMed

Yang, Yu; Zhou, Xiaofeng; Gao, Shuangqing; Lin, Hongbo; Xie, Yanming; Feng, Yuji; Huang, Kui; Zhan, Siyan

2018-01-01

Electronic healthcare databases (EHDs) are used increasingly for post-marketing drug safety surveillance and pharmacoepidemiology in Europe and North America. However, few studies have examined the potential of these data sources in China. Three major types of EHDs in China (i.e., a regional community-based database, a national claims database, and an electronic medical records [EMR] database) were selected for evaluation. Forty core variables were derived based on the US Mini-Sentinel (MS) Common Data Model (CDM) as well as the data features in China that would be desirable to support drug safety surveillance. An email survey of these core variables and eight general questions as well as follow-up inquiries on additional variables was conducted. These 40 core variables across the three EHDs and all variables in each EHD along with those in the US MS CDM and Observational Medical Outcomes Partnership (OMOP) CDM were compared for availability and labeled based on specific standards. All of the EHDs' custodians confirmed their willingness to share their databases with academic institutions after appropriate approval was obtained. The regional community-based database contained 1.19 million people in 2015 with 85% of core variables. Resampled annually nationwide, the national claims database included 5.4 million people in 2014 with 55% of core variables, and the EMR database included 3 million inpatients from 60 hospitals in 2015 with 80% of core variables. Compared with MS CDM or OMOP CDM, the proportion of variables across the three EHDs available or able to be transformed/derived from the original sources are 24-83% or 45-73%, respectively. These EHDs provide potential value to post-marketing drug safety surveillance and pharmacoepidemiology in China. Future research is warranted to assess the quality and completeness of these EHDs or additional data sources in China.

Patient safety education at Japanese medical schools: results of a nationwide survey

PubMed Central

2012-01-01

Background Patient safety education, including error prevention strategies and management of adverse events, has become a topic of worldwide concern. The importance of the patient safety is also recognized in Japan following two serious medical accidents in 1999. Furthermore, educational curriculum guideline revisions in 2008 by relevant the Ministry of Education includes patient safety as part of the core medical curriculum. However, little is known about the patient safety education in Japanese medical schools partly because a comprehensive study has not yet been conducted in this field. Therefore, we have conducted a nationwide survey in order to clarify the current status of patient safety education at medical schools in Japan. Results Response rate was 60.0% (n = 48/80). Ninety-eight-percent of respondents (n = 47/48) reported integration of patient safety education into their curricula. Thirty-nine percent reported devoting less than five hours to the topic. All schools that teach patient safety reported use of lecture based teaching methods while few used alternative methods, such as role-playing or in-hospital training. Topics related to medical error theory and legal ramifications of error are widely taught while practical topics related to error analysis such as root cause analysis are less often covered. Conclusions Based on responses to our survey, most Japanese medical schools have incorporated the topic of patient safety into their curricula. However, the number of hours devoted to the patient safety education is far from the sufficient level with forty percent of medical schools that devote five hours or less to it. In addition, most medical schools employ only the lecture based learning, lacking diversity in teaching methods. Although most medical schools cover basic error theory, error analysis is taught at fewer schools. We still need to make improvements to our medical safety curricula. We believe that this study has the implications for the rest of the world as a model of what is possible and a sounding board for what topics might be important. PMID:22574712

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

A Human Factors Analysis and Classification System (HFACS) Examination of Commercial Vessel Accidents

DTIC Science & Technology

2012-09-01

Naval Operations before the Congress on FY2013 Department of Navy posture. Heinrich , H . W. (1941). Industrial accident prevention : A scientific...Theory The core of the Domino Theory, developed by Herbert W. Heinrich who studied industrial safety in the early 1900s, is that accidents are a result...chain of events resulting in an accident . Heinrich likened the dominos to unsafe conditions or unsafe acts, where their subsequent removal prevents a

Mutagenicity in a Molecule: Identification of Core Structural Features of Mutagenicity Using a Scaffold Analysis

PubMed Central

Hsu, Kuo-Hsiang; Su, Bo-Han; Tu, Yi-Shu; Lin, Olivia A.; Tseng, Yufeng J.

2016-01-01

With advances in the development and application of Ames mutagenicity in silico prediction tools, the International Conference on Harmonisation (ICH) has amended its M7 guideline to reflect the use of such prediction models for the detection of mutagenic activity in early drug safety evaluation processes. Since current Ames mutagenicity prediction tools only focus on functional group alerts or side chain modifications of an analog series, these tools are unable to identify mutagenicity derived from core structures or specific scaffolds of a compound. In this study, a large collection of 6512 compounds are used to perform scaffold tree analysis. By relating different scaffolds on constructed scaffold trees with Ames mutagenicity, four major and one minor novel mutagenic groups of scaffold are identified. The recognized mutagenic groups of scaffold can serve as a guide for medicinal chemists to prevent the development of potentially mutagenic therapeutic agents in early drug design or development phases, by modifying the core structures of mutagenic compounds to form non-mutagenic compounds. In addition, five series of substructures are provided as recommendations, for direct modification of potentially mutagenic scaffolds to decrease associated mutagenic activities. PMID:26863515

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

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

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Developing a patient-led electronic feedback system for quality and safety within Renal PatientView.

PubMed

Giles, Sally J; Reynolds, Caroline; Heyhoe, Jane; Armitage, Gerry

2017-03-01

It is increasingly acknowledged that patients can provide direct feedback about the quality and safety of their care through patient reporting systems. The aim of this study was to explore the feasibility of patients, healthcare professionals and researchers working in partnership to develop a patient-led quality and safety feedback system within an existing electronic health record (EHR), known as Renal PatientView (RPV). Phase 1 (inception) involved focus groups (n = 9) and phase 2 (requirements) involved cognitive walkthroughs (n = 34) and 1:1 qualitative interviews (n = 34) with patients and healthcare professionals. A Joint Services Expert Panel (JSP) was convened to review the findings from phase 1 and agree the core principles and components of the system prototype. Phase 1 data were analysed using a thematic approach. Data from phase 1 were used to inform the design of the initial system prototype. Phase 2 data were analysed using the components of heuristic evaluation, resulting in a list of core principles and components for the final system prototype. Phase 1 identified four main barriers and facilitators to patients feeding back on quality and safety concerns. In phase 2, the JSP agreed that the system should be based on seven core principles and components. Stakeholders were able to work together to identify core principles and components for an electronic patient quality and safety feedback system in renal services. Tensions arose due to competing priorities, particularly around anonymity and feedback. Careful consideration should be given to the feasibility of integrating a novel element with differing priorities into an established system with existing functions and objectives. © 2016 European Dialysis and Transplant Nurses Association/European Renal Care Association.

Foundational workplace safety and health competencies for the emerging workforce.

PubMed

Okun, Andrea H; Guerin, Rebecca J; Schulte, Paul A

2016-12-01

Young workers (aged 15-24) suffer disproportionately from workplace injuries, with a nonfatal injury rate estimated to be two times higher than among workers age 25 or over. These workers make up approximately 9% of the U.S. workforce and studies have shown that nearly 80% of high school students work at some point during high school. Although young worker injuries are a pressing public health problem, the critical knowledge and skills needed to prepare youth for safe and healthy work are missing from most frameworks used to prepare the emerging U.S. workforce. A framework of foundational workplace safety and health knowledge and skills (the NIOSH 8 Core Competencies) was developed based on the Health Belief Model (HBM). The proposed NIOSH Core Competencies utilize the HBM to provide a framework for foundational workplace safety and health knowledge and skills. An examination of how these competencies and the HBM apply to actions that workers take to protect themselves is provided. The social and physical environments that influence these actions are also discussed. The NIOSH 8 Core Competencies, grounded in one of the most widely used health behavior theories, fill a critical gap in preparing the emerging U.S. workforce to be cognizant of workplace risks. Integration of the NIOSH 8 Core Competencies into school curricula is one way to ensure that every young person has the foundational workplace safety and health knowledge and skills to participate in, and benefit from, safe and healthy work. Published by Elsevier Ltd.

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

The Core Values that Support Health, Safety, and Well-being at Work

PubMed Central

Zwetsloot, Gerard I.J.M.; Scheppingen, Arjella R. van; Bos, Evelien H.; Dijkman, Anja; Starren, Annick

2013-01-01

Background Health, safety, and well-being (HSW) at work represent important values in themselves. It seems, however, that other values can contribute to HSW. This is to some extent reflected in the scientific literature in the attention paid to values like trust or justice. However, an overview of what values are important for HSW was not available. Our central research question was: what organizational values are supportive of health, safety, and well-being at work? Methods The literature was explored via the snowball approach to identify values and value-laden factors that support HSW. Twenty-nine factors were identified as relevant, including synonyms. In the next step, these were clustered around seven core values. Finally, these core values were structured into three main clusters. Results The first value cluster is characterized by a positive attitude toward people and their “being”; it comprises the core values of interconnectedness, participation, and trust. The second value cluster is relevant for the organizational and individual “doing”, for actions planned or undertaken, and comprises justice and responsibility. The third value cluster is relevant for “becoming” and is characterized by the alignment of personal and organizational development; it comprises the values of growth and resilience. Conclusion The three clusters of core values identified can be regarded as “basic value assumptions” that underlie both organizational culture and prevention culture. The core values identified form a natural and perhaps necessary aspect of a prevention culture, complementary to the focus on rational and informed behavior when dealing with HSW risks. PMID:24422174

Performance of food safety management systems in poultry meat preparation processing plants in relation to Campylobacter spp. contamination.

PubMed

Sampers, Imca; Jacxsens, Liesbeth; Luning, Pieternel A; Marcelis, Willem J; Dumoulin, Ann; Uyttendaele, Mieke

2010-08-01

A diagnostic instrument comprising a combined assessment of core control and assurance activities and a microbial assessment instrument were used to measure the performance of current food safety management systems (FSMSs) of two poultry meat preparation companies. The high risk status of the company's contextual factors, i.e., starting from raw materials (poultry carcasses) with possible high numbers and prevalence of pathogens such as Campylobacter spp., requires advanced core control and assurance activities in the FSMS to guarantee food safety. The level of the core FSMS activities differed between the companies, and this difference was reflected in overall microbial quality (mesophilic aerobic count), presence of hygiene indicators (Enterobacteriaceae, Staphylococcus aureus, and Escherichia coli), and contamination with pathogens such as Salmonella, Listeria monocytogenes, and Campylobacter spp. The food safety output expressed as a microbial safety profile was related to the variability in the prevalence and contamination levels of Campylobacter spp. in poultry meat preparations found in a Belgian nationwide study. Although a poultry meat processing company could have an advanced FSMS in place and a good microbial profile (i.e., lower prevalence of pathogens, lower microbial numbers, and less variability in microbial contamination), these positive factors might not guarantee pathogen-free products. Contamination could be attributed to the inability to apply effective interventions to reduce or eliminate pathogens in the production chain of (raw) poultry meat preparations.

Root causes and impacts of severe accidents at large nuclear power plants.

PubMed

Högberg, Lars

2013-04-01

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

Public Awareness Survey Recommendations of the NHTSA-GHSA Working Group

DOT National Transportation Integrated Search

2011-07-01

The Governors Highway Safety Association (GHSA) and the National Highway Traffic Safety Administration (NHTSA) developed a basic set of survey questions including information on seat belt use, impaired driving, and speeding. These core questions can ...

Core-shell structured ceramic nonwoven separators by atomic layer deposition for safe lithium-ion batteries

NASA Astrophysics Data System (ADS)

Shen, Xiu; Li, Chao; Shi, Chuan; Yang, Chaochao; Deng, Lei; Zhang, Wei; Peng, Longqing; Dai, Jianhui; Wu, Dezhi; Zhang, Peng; Zhao, Jinbao

2018-05-01

Safety is one of the most factors for lithium-ion batteries (LIBs). In this work, a novel kind of ceramic separator with high safety insurance is proposed. We fabricated the core-shell nanofiber separators for LIBs by atomic layer deposition (ALD) of 30 nm Al2O3 on the electrospinning nonwoven fiber of polyvinylidene fluoride-hexafluoropropylene (PVDF-HFP). The separators show a pretty high heat resistance up to 200 °C without any shrinkage, an excellent fire-resistant property and a wide electrochemical window. Besides, with higher uptake and ionic conductivity, cells assembled with the novel separator shows better electrochemical performance. The ALD produced separators exhibit great potential in elaborate products like 3C communications and in energy field with harsh requirements for safety such as electric vehicles. The application of ALD on polymer fiber membranes brings a new strategy and opportunity for improving the safety of the advanced LIBs.

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)

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

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

Kraus, A.; Hu, R.

2015-01-01

Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves overmore » the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.« less

Training infection control and hospital hygiene professionals in Europe, 2010: agreed core competencies among 33 European countries.

PubMed

Brusaferro, S; Cookson, B; Kalenic, S; Cooper, T; Fabry, J; Gallagher, R; Hartemann, P; Mannerquist, K; Popp, W; Privitera, G; Ruef, C; Viale, P; Coiz, F; Fabbro, E; Suetens, C; Varela Santos, C

2014-12-11

The harmonisation of training programmes for infection control and hospital hygiene (IC/HH) professionals in Europe is a requirement of the Council recommendation on patient safety. The European Centre for Disease Prevention and Control commissioned the 'Training Infection Control in Europe' project to develop a consensus on core competencies for IC/HH professionals in the European Union (EU). Core competencies were drafted on the basis of the Improving Patient Safety in Europe (IPSE) project's core curriculum (CC), evaluated by questionnaire and approved by National Representatives (NRs) for IC/HH training. NRs also re-assessed the status of IC/HH training in European countries in 2010 in comparison with the situation before the IPSE CC in 2006. The IPSE CC had been used to develop or update 28 of 51 IC/HH courses. Only 10 of 33 countries offered training and qualification for IC/HH doctors and nurses. The proposed core competencies are structured in four areas and 16 professional tasks at junior and senior level. They form a reference for standardisation of IC/HH professional competencies and support recognition of training initiatives.

Analysis of Counterfeit Coated Tablets and Multi-Layer Packaging Materials Using Infrared Microspectroscopic Imaging.

PubMed

Winner, Taryn L; Lanzarotta, Adam; Sommer, André J

2016-06-01

An effective method for detecting and characterizing counterfeit finished dosage forms and packaging materials is described in this study. Using attenuated total internal reflection Fourier transform infrared spectroscopic imaging, suspect tablet coating and core formulations as well as multi-layered foil safety seals, bottle labels, and cigarette tear tapes were analyzed and compared directly with those of a stored authentic product. The approach was effective for obtaining molecular information from structures as small as 6 μm.

LOGIC CIRCUIT

DOEpatents

Strong, G.H.; Faught, M.L.

1963-12-24

A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

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).

Understanding the haling power depletion (HPD) method

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

Levine, S.; Blyth, T.; Ivanov, K.

2012-07-01

The Pennsylvania State Univ. (PSU) is using the university version of the Studsvik Scandpower Code System (CMS) for research and education purposes. Preparations have been made to incorporate the CMS into the PSU Nuclear Engineering graduate class 'Nuclear Fuel Management' course. The information presented in this paper was developed during the preparation of the material for the course. The Haling Power Depletion (HPD) was presented in the course for the first time. The HPD method has been criticized as not valid by many in the field even though it has been successfully applied at PSU for the past 20 years.more » It was noticed that the radial power distribution (RPD) for low leakage cores during depletion remained similar to that of the HPD during most of the cycle. Thus, the Haling Power Depletion (HPD) may be used conveniently mainly for low leakage cores. Studies were then made to better understand the HPD and the results are presented in this paper. Many different core configurations can be computed quickly with the HPD without using Burnable Poisons (BP) to produce several excellent low leakage core configurations that are viable for power production. Once the HPD core configuration is chosen for further analysis, techniques are available for establishing the BP design to prevent violating any of the safety constraints in such HPD calculated cores. In summary, in this paper it has been shown that the HPD method can be used for guiding the design for the low leakage core. (authors)« less

Advancing interprofessional patient safety education for medical, nursing, and pharmacy learners during clinical rotations.

PubMed

Thom, Kerri A; Heil, Emily L; Croft, Lindsay D; Duffy, Alison; Morgan, Daniel J; Johantgen, Mary

2016-11-01

Clinical errors are common and can lead to adverse events and patient death. Health professionals must work within interprofessional teams to provide safe and effective care to patients, yet current curricula is lacking with regards to interprofessional education and patient safety. We describe the development and implementation of an interprofessional course aimed at medical, nursing, and pharmacy learners during their clinical training at a large academic medical centre. The course objectives were based on core competencies for interprofessional education and patient safety. The course was offered as recurring three 1-hour sessions, including case-based discussions and a mock root cause analysis. Forty-three students attended at least one session over a 7-month period. We performed a cross-sectional survey of participants to assess readiness for interprofessional learning and a before and after comparison of patient safety knowledge. All students reported a high level of readiness for interprofessional learning, indicating an interest in interprofessional opportunities. In general, understanding and knowledge of the four competency domains in patient safety was low before the course and 100% of students reported an increase in knowledge in these domains after participating in the course.

Factors impacting perceived safety among staff working on mental health wards

PubMed Central

Brown, Andrew; McCabe, Rhiannah; Rogerson, Michelle; Whittington, Richard

2017-01-01

Background Safety at work is a core issue for mental health staff working on in-patient units. At present, there is a limited theoretical base regarding which factors may affect staff perceptions of safety. Aims This study attempted to identify which factors affect perceived staff safety working on in-patient mental health wards. Method A cross-sectional design was employed across 101 forensic and non-forensic mental health wards, over seven National Health Service trusts nationally. Measures included an online staff survey, Ward Features Checklist and recorded incident data. Data were analysed using categorical principal components analysis and ordinal regression. Results Perceptions of staff safety were increased by ward brightness, higher number of patient beds, lower staff to patient ratios, less dayroom space and more urban views. Conclusions The findings from this study do not represent common-sense assumptions. Results are discussed in the context of the literature and may have implications for current initiatives aimed at managing in-patient violence and aggression. Declaration of interest None. Copyright and usage © The Royal College of Psychiatrists 2017. This is an open access article distributed under the terms of the Creative Commons Attribution (CC BY) license. PMID:28904814

A NEW METHOD TO QUANTIFY CORE TEMPERATURE INSTABILITY IN RODENTS.

EPA Science Inventory

Methods to quantify instability of autonomic systems such as temperature regulation should be important in toxicant and drug safety studies. Stability of core temperature (Tc) in laboratory rodents is susceptible to a variety of stimuli. Calculating the temperature differential o...

Laboratory ultrasonic pulse velocity logging for determination of elastic properties from rock core

NASA Astrophysics Data System (ADS)

Blacklock, Natalie Erin

During the development of deep underground excavations spalling and rockbursting have been recognized as significant mechanisms of violent brittle failure. In order to predict whether violent brittle failure will occur, it is important to identify the location of stiffness transitions that are associated with geologic structure. One approach to identify the effect of geologic structures is to apply borehole geophysical tools ahead of the tunnel advance. Stiffness transitions can be identified using mechanical property analysis surveys that combine acoustic velocity and density data to calculate acoustic estimates of elastic moduli. However, logistical concerns arise since the approach must be conducted at the advancing tunnel face. As a result, borehole mechanical property analyses are rarely used. Within this context, laboratory ultrasonic pulse velocity testing has been proposed as a potential alternative to borehole mechanical property analysis since moving the analysis to the laboratory would remove logistical constraints and improve safety for the evaluators. In addition to the traditional method of conducting velocity testing along the core axis, two new methodologies for point-focused testing were developed across the core diameter, and indirectly along intact lengths of drill core. The indirect test procedure was implemented in a continuous ultrasonic velocity test program along 573m of drill core to identify key geologic structures that generated transitions in ultrasonic elastic moduli. The test program was successful at identifying the location of geologic contacts, igneous intrusions, faults and shear structures. Ultrasonic values of Young's modulus and bulk modulus were determined at locations of significant velocity transitions to examine the potential for energy storage and energy release. Comparison of results from different ultrasonic velocity test configurations determined that the indirect test configuration provided underestimates for values of Young's modulus. This indicated that the test procedure will require modifications to improve coupling of the transducers to the core surface. In order to assess whether laboratory testing can be an alternative to borehole surveys, laboratory velocity testing must be directly assessed with results from acoustic borehole logging. There is also potential for the laboratory velocity program to be used to assess small scale stiffness changes, differences in mineral composition and the degree of fracturing of drill core.

John M. Eisenberg Patient Safety Awards. System innovation: Veterans Health Administration National Center for Patient Safety.

PubMed

Heget, Jeffrey R; Bagian, James P; Lee, Caryl Z; Gosbee, John W

2002-12-01

In 1998 the Veterans Health Administration (VHA) created the National Center for Patient Safety (NCPS) to lead the effort to reduce adverse events and close calls systemwide. NCPS's aim is to foster a culture of safety in the Department of Veterans Affairs (VA) by developing and providing patient safety programs and delivering standardized tools, methods, and initiatives to the 163 VA facilities. To create a system-oriented approach to patient safety, NCPS looked for models in fields such as aviation, nuclear power, human factors, and safety engineering. Core concepts included a non-punitive approach to patient safety activities that emphasizes systems-based learning, the active seeking out of close calls, which are viewed as opportunities for learning and investigation, and the use of interdisciplinary teams to investigate close calls and adverse events through a root cause analysis (RCA) process. Participation by VA facilities and networks was voluntary. NCPS has always aimed to develop a program that would be applicable both within the VA and beyond. NCPS's full patient safety program was tested and implemented throughout the VA system from November 1999 to August 2000. Program components included an RCA system for use by caregivers at the front line, a system for the aggregate review of RCA results, information systems software, alerts and advisories, and cognitive acids. Following program implementation, NCPS saw a 900-fold increase in reporting of close calls of high-priority events, reflecting the level of commitment to the program by VHA leaders and staff.

Deployment of a tool for measuring freeway safety performance.

DOT National Transportation Integrated Search

2011-12-01

This project updated and deployed a freeway safety performance measurement tool, building upon a previous project that developed the core methodology. The tool evaluates the cumulative risk over time of an accident or a particular kind of accident. T...

Public awareness survey recommendations of the NHTSA-GHSA working group : traffic tech.

DOT National Transportation Integrated Search

2010-10-01

The National Highway Transportation Safety Administration : and the Governors Highway Safety Association : (GHSA) developed a set of survey questions about seat : belt use, impaired driving, and speeding. Using the : same core questions in all data c...

A study of passive safety features by utilizing intra-subassembly-equipped self-actuated shutdown mechanism for future large fast breeder reactors

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

Uto, N.; Niwa, H.; Ieda, Y.

1996-08-01

Passive prevention of core disruptive accidents (CDAs) is desired in terms of enhancement of safety for future fast breeder reactors. In addition, mitigation of CDA`s consequences should be required because mitigation measures have a potential of applying to all accidents, while prevention measures are prepared for specific accident initiators. In this paper, the Intra-Subassembly-equipped Self-Actuated Shutdown System (IS-SASS) , which is considered effective on passive prevention and mitigation of CDAs, is described. The IS-SASS is introduced in a fuel subassembly and consists of absorber materials at the top of the active core and an inner duct through which molten fuelmore » can be excluded out of the core. The determination of the appropriate number of the IS-SASS units, their arrangement in the core and their suitable structure are found to be suited to prevention and mitigation of CDAs for liquid metal-cooled large fast breeder reactors.« less

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

Naval War College Review. Volume 64, Number 2, Spring 2011

DTIC Science & Technology

2011-01-01

to revolutionize the African maritime sector holistically, across its entire spectrum—improving safety and security, gover - nance, and industrial...strategy for a maritime economy that includes the enabling elements of gover - nance, infrastructure, trade, safety, and security and plainly tells global...transport in its core function); tourism ; energy; infrastructure (ports); cooperation on safety, security, and environmental protection; tariff harmonization

Thermal stress in North Western Australian iron ore mining staff.

PubMed

Peiffer, Jeremiah J; Abbiss, Chris R

2013-05-01

Demand for Australian mined iron ore has increased employment within this sector, thus exposing increased numbers of workers to the harsh Australian climate. This study examined the influence of hot (>30°C wet bulb globe temperature) environmental temperatures, consistent with working in North Western Australia, on iron ore mining staff. Core temperature, hydration status, perceived exertion, mood, and fatigue state were measured in 77 participants at three time points (pre-, mid-, and post-shift) during a normal 12-h shift at an open-cut iron ore mining/processing site (n = 31; Site1) and an iron ore processing/shipping site (n = 46; Site2). A significant effect for time was observed for core temperature with greater mean core temperatures measured mid-shift (37.5±0.4°C) and post-shift (37.6±0.3°C) compared with pre-shift values (37.0±0.5°C). All mean core temperature measures were lower than ISO7933 thresholds (38°C) for thermal safety. Mean hydration measures [urine-specific gravity (USG)] were greater at Site1 (1.029±0.006) compared with those at Site2 (1.021±0.007). Furthermore, both pre- and post-shift measures from Site1 and the post-shift measures from Site2 were greater than the threshold for dehydration (USG = 1.020). No differences were observed for mood or perceived exertion over time; however, measures of fatigue state were greater post-shift compared with pre- and mid-shift values for both sites. Our findings indicate that the majority of mine workers in North Western Australia are able to regulate work rate in hot environments to maintain core temperatures below ISO safety guidelines; however, 22% of workers reached or exceeded the safety guidelines, warranting further investigation. Furthermore, hydration practices, especially when off-work, appear inadequate and could endanger health and safety.

Transcatheter aortic valve replacement using a self-expanding bioprosthesis in patients with severe aortic stenosis at extreme risk for surgery.

PubMed

Popma, Jeffrey J; Adams, David H; Reardon, Michael J; Yakubov, Steven J; Kleiman, Neal S; Heimansohn, David; Hermiller, James; Hughes, G Chad; Harrison, J Kevin; Coselli, Joseph; Diez, Jose; Kafi, Ali; Schreiber, Theodore; Gleason, Thomas G; Conte, John; Buchbinder, Maurice; Deeb, G Michael; Carabello, Blasé; Serruys, Patrick W; Chenoweth, Sharla; Oh, Jae K

2014-05-20

This study sought to evaluate the safety and efficacy of the CoreValve transcatheter heart valve (THV) for the treatment of severe aortic stenosis in patients at extreme risk for surgery. Untreated severe aortic stenosis is a progressive disease with a poor prognosis. Transcatheter aortic valve replacement (TAVR) with a self-expanding bioprosthesis is a potentially effective therapy. We performed a prospective, multicenter, nonrandomized investigation evaluating the safety and efficacy of self-expanding TAVR in patients with symptomatic severe aortic stenosis with prohibitive risks for surgery. The primary endpoint was a composite of all-cause mortality or major stroke at 12 months, which was compared with a pre-specified objective performance goal (OPG). A total of 41 sites in the United States recruited 506 patients, of whom 489 underwent attempted treatment with the CoreValve THV. The rate of all-cause mortality or major stroke at 12 months was 26.0% (upper 2-sided 95% confidence bound: 29.9%) versus 43.0% with the OPG (p < 0.0001). Individual 30-day and 12-month events included all-cause mortality (8.4% and 24.3%, respectively) and major stroke (2.3% and 4.3%, respectively). Procedural events at 30 days included life-threatening/disabling bleeding (12.7%), major vascular complications (8.2%), and need for permanent pacemaker placement (21.6%). The frequency of moderate or severe paravalvular aortic regurgitation was lower 12 months after self-expanding TAVR (4.2%) than at discharge (10.7%; p = 0.004 for paired analysis). TAVR with a self-expanding bioprosthesis was safe and effective in patients with symptomatic severe aortic stenosis at prohibitive risk for surgical valve replacement. (Safety and Efficacy Study of the Medtronic CoreValve System in the Treatment of Symptomatic Severe Aortic Stenosis in High Risk and Very High Risk Subjects Who Need Aortic Valve Replacement; NCT01240902). Copyright © 2014 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

Wet--But Safe. A Classroom Course in Water Safety and Survival.

ERIC Educational Resources Information Center

Michigan State Dept. of Natural Resources, Lansing.

This manual is designed for use in elementary school systems that do not have a swimming pool available. It contains eight classroom sessions and provides a core of information in basic water safety, water survival, and water rescue. (JD)

Evaluating the Influence of Nutrition Determinants on Construction Workers' Food Choices.

PubMed

Okoro, Chioma Sylvia; Musonda, Innocent; Agumba, Justus

2017-11-01

Nutritional knowledge as well as economic, social, biological, and cultural factors have been known to determine an individual's food choices. Despite the existence of research on the factors which influence nutrition globally, there is little known about the extent to which these factors influence the food choices of construction workers, which in turn influence their health and safety during construction activities. The present article investigates the extent to which construction workers' nutrition is influenced by nutritional knowledge, as well as economic, environmental, social, psychological, and physiological factors. A field questionnaire survey was conducted on site construction workers in the Gauteng Province of South Africa. Principal components analysis and multiple regression analysis were used to analyze the data. Findings revealed that consumption of foods termed alternative foods including dairy products, eggs, nuts, fish, and cereals, was influenced by nutritional knowledge and resources. Foods termed traditional core foods were influenced by cultural background; foods termed secondary core foods comprising fruits and vegetables were influenced by economic factors, resources, and cultural background; while foods termed core foods were mostly influenced by nutritional knowledge. By providing evidence of the factors which most influence selection and consumption of certain foods by construction workers, relevant nutrition interventions will be designed and implemented, taking cognizance of these factors.

Evaluating the Influence of Nutrition Determinants on Construction Workers’ Food Choices

PubMed Central

Okoro, Chioma Sylvia; Musonda, Innocent; Agumba, Justus

2016-01-01

Nutritional knowledge as well as economic, social, biological, and cultural factors have been known to determine an individual’s food choices. Despite the existence of research on the factors which influence nutrition globally, there is little known about the extent to which these factors influence the food choices of construction workers, which in turn influence their health and safety during construction activities. The present article investigates the extent to which construction workers’ nutrition is influenced by nutritional knowledge, as well as economic, environmental, social, psychological, and physiological factors. A field questionnaire survey was conducted on site construction workers in the Gauteng Province of South Africa. Principal components analysis and multiple regression analysis were used to analyze the data. Findings revealed that consumption of foods termed alternative foods including dairy products, eggs, nuts, fish, and cereals, was influenced by nutritional knowledge and resources. Foods termed traditional core foods were influenced by cultural background; foods termed secondary core foods comprising fruits and vegetables were influenced by economic factors, resources, and cultural background; while foods termed core foods were mostly influenced by nutritional knowledge. By providing evidence of the factors which most influence selection and consumption of certain foods by construction workers, relevant nutrition interventions will be designed and implemented, taking cognizance of these factors. PMID:26821794

Human factors and ergonomics as a patient safety practice

PubMed Central

Carayon, Pascale; Xie, Anping; Kianfar, Sarah

2014-01-01

Background Human factors and ergonomics (HFE) approaches to patient safety have addressed five different domains: usability of technology; human error and its role in patient safety; the role of healthcare worker performance in patient safety; system resilience; and HFE systems approaches to patient safety. Methods A review of various HFE approaches to patient safety and studies on HFE interventions was conducted. Results This paper describes specific examples of HFE-based interventions for patient safety. Studies show that HFE can be used in a variety of domains. Conclusions HFE is a core element of patient safety improvement. Therefore, every effort should be made to support HFE applications in patient safety. PMID:23813211

A hybrid simulation approach for integrating safety behavior into construction planning: An earthmoving case study.

PubMed

Goh, Yang Miang; Askar Ali, Mohamed Jawad

2016-08-01

One of the key challenges in improving construction safety and health is the management of safety behavior. From a system point of view, workers work unsafely due to system level issues such as poor safety culture, excessive production pressure, inadequate allocation of resources and time and lack of training. These systemic issues should be eradicated or minimized during planning. However, there is a lack of detailed planning tools to help managers assess the impact of their upstream decisions on worker safety behavior. Even though simulation had been used in construction planning, the review conducted in this study showed that construction safety management research had not been exploiting the potential of simulation techniques. Thus, a hybrid simulation framework is proposed to facilitate integration of safety management considerations into construction activity simulation. The hybrid framework consists of discrete event simulation (DES) as the core, but heterogeneous, interactive and intelligent (able to make decisions) agents replace traditional entities and resources. In addition, some of the cognitive processes and physiological aspects of agents are captured using system dynamics (SD) approach. The combination of DES, agent-based simulation (ABS) and SD allows a more "natural" representation of the complex dynamics in construction activities. The proposed hybrid framework was demonstrated using a hypothetical case study. In addition, due to the lack of application of factorial experiment approach in safety management simulation, the case study demonstrated sensitivity analysis and factorial experiment to guide future research. Copyright © 2015 Elsevier Ltd. All rights reserved.

Fundamentally updating fundamentals.

PubMed

Armstrong, Gail; Barton, Amy

2013-01-01

Recent educational research indicates that the six competencies of the Quality and Safety Education for Nurses initiative are best introduced in early prelicensure clinical courses. Content specific to quality and safety has traditionally been covered in senior level courses. This article illustrates an effective approach to using quality and safety as an organizing framework for any prelicensure fundamentals of nursing course. Providing prelicensure students a strong foundation in quality and safety in an introductory clinical course facilitates early adoption of quality and safety competencies as core practice values. Copyright © 2013 Elsevier Inc. All rights reserved.

Temperature actuated automatic safety rod release

DOEpatents

Hutter, E.; Pardini, J.A.; Walker, D.E.

1984-03-13

A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

Temperature actuated automatic safety rod release

DOEpatents

Hutter, Ernest; Pardini, John A.; Walker, David E.

1987-01-01

A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

An improved tri-tube cryogenic gravel sampler.

Treesearch

Fred H. Everest; Carl E. McLemore; John F. Ward

1980-01-01

The tri-tube cryogenic gravel sampler has been improved, and accessories have been developed that increase its reliability and safety of operation, reduce core extraction time, and allow accurate partitioning of cores into subsamples. The improved tri-tube sampler is one of the most versatile and efficient substrate sampling tools yet developed.

Two Year Core Curriculum for Agricultural Education in Montana. Revised.

ERIC Educational Resources Information Center

Montana State Univ., Bozeman. Dept. of Agricultural and Industrial Education.

This core curriculum consists of materials for use in conducting a two-year secondary level agricultural education course. Addressed in the individual units of the guide are the following topics: leadership; agricultural career planning; supervised occupational experience programs (SOEPs); agricultural mechanics (shop management and safety,…

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

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

David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Updatemore » Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.« less

A PVDF-Based Sensor for Internal Stress Monitoring of a Concrete-Filled Steel Tubular (CFST) Column Subject to Impact Loads.

PubMed

Du, Guofeng; Li, Zhao; Song, Gangbing

2018-05-23

Impact loads can have major adverse effects on the safety of civil engineering structures, such as concrete-filled steel tubular (CFST) columns. The study of mechanical behavior and stress analysis of CFST columns under impact loads is very important to ensure their safety against such loads. At present, the internal stress monitoring of the concrete cores CFST columns under impact loads is still a very challenging subject. In this paper, a PVDF (Polyvinylidene Fluoride) piezoelectric smart sensor was developed and successfully applied to the monitoring of the internal stress of the concrete core of a CFST column under impact loads. The smart sensor consists of a PVDF piezoelectric film sandwiched between two thin steel plates through epoxy. The protection not only prevents the PVDF film from impact damages but also ensures insulation and waterproofing. The smart sensors were embedded into the circular concrete-filled steel tube specimen during concrete pouring. The specimen was tested against impact loads, and testing data were collected. The time history of the stress obtained from the PVDF smart sensor revealed the evolution of core concrete internal stress under impact loads when compared with the impact force⁻time curve of the hammer. Nonlinear finite element simulations of the impact process were also carried out. The results of FEM simulations had good agreement with the test results. The results showed that the proposed PVDF piezoelectric smart sensors can effectively monitor the internal stress of concrete-filled steel tubular columns under impact loads.

Tailoring the response of Autonomous Reactivity Control (ARC) systems

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

Qvist, Staffan A.; Hellesen, Carl; Gradecka, Malwina

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, weremore » performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.« less

Multi-sectoral action for child safety-a European study exploring implicated sectors.

PubMed

Scholtes, Beatrice; Schröder-Bäck, Peter; Förster, Katharina; MacKay, Morag; Vincenten, Joanne; Brand, Helmut

2017-06-01

Injury to children in Europe, resulting in both death and disability, constitutes a significant burden on individuals, families and society. Inequalities between high and low-income countries are growing. The World Health Organisation Health 2020 strategy calls for inter-sectoral collaboration to address injury in Europe and advocates the whole of government and whole of society approaches to wicked problems. In this study we explore which sectors (e.g. health, transport, education) are relevant for four domains of child safety (intentional injury, water, road and home safety). We used the organigraph methodology, originally developed to demonstrate how organizations work, to describe the governance of child safety interventions. Members of the European Child Safety Alliance, working in the field of child safety in 24 European countries, drew organigraphs of evidence-based interventions. They included the different actors involved and the processes between them. We analyzed the organigraphs by counting the actors presented and categorizing them into sectors using a pre-defined analysis framework. We received 44 organigraphs from participants in 24 countries. Twenty-seven sectors were identified across the four domains. Nine of the 27 identified sectors were classified as 'core sectors' (education, health, home affairs, justice, media, recreation, research, social/welfare services and consumers). This study reveals the multi-sectoral nature of child safety in practice. It provides information for stakeholders working in child safety to help them implement inter-sectoral child safety interventions taking a whole-of-government and whole-of-society approach to health governance. © The Author 2017. Published by Oxford University Press on behalf of the European Public Health Association. All rights reserved.

Evidence-based safety (EBS) management: A new approach to teaching the practice of safety management (SM).

PubMed

Wang, Bing; Wu, Chao; Shi, Bo; Huang, Lang

2017-12-01

In safety management (SM), it is important to make an effective safety decision based on the reliable and sufficient safety-related information. However, many SM failures in organizations occur for a lack of the necessary safety-related information for safety decision-making. Since facts are the important basis and foundation for decision-making, more efforts to seek the best evidence relevant to a particular SM problem would lead to a more effective SM solution. Therefore, the new paradigm for decision-making named "evidence-based practice (EBP)" can hold important implications for SM, because it uses the current best evidence for effective decision-making. Based on a systematic review of existing SM approaches and an analysis of reasons why we need new SM approaches, we created a new SM approach called evidence-based safety (EBS) management by introducing evidence-based practice into SM. It was necessary to create new SM approaches. A new SM approach called EBS was put forward, and the basic questions of EBS such as its definition and core were analyzed in detail. Moreover, the determinants of EBS included manager's attitudes towards EBS; evidence-based consciousness in SM; evidence sources; technical support; EBS human resources; organizational culture; and individual attributes. EBS is a new and effective approach to teaching the practice of SM. Of course, further research on EBS should be carried out to make EBS a reality. Practical applications: Our work can provide a new and effective idea and method to teach the practice of SM. Specifically, EBS proposed in our study can help safety professionals make an effective safety decision based on a firm foundation of high-grade evidence. Copyright © 2017 National Safety Council and Elsevier Ltd. All rights reserved.

Benchmarking road safety performance: Identifying a meaningful reference (best-in-class).

PubMed

Chen, Faan; Wu, Jiaorong; Chen, Xiaohong; Wang, Jianjun; Wang, Di

2016-01-01

For road safety improvement, comparing and benchmarking performance are widely advocated as the emerging and preferred approaches. However, there is currently no universally agreed upon approach for the process of road safety benchmarking, and performing the practice successfully is by no means easy. This is especially true for the two core activities of which: (1) developing a set of road safety performance indicators (SPIs) and combining them into a composite index; and (2) identifying a meaningful reference (best-in-class), one which has already obtained outstanding road safety practices. To this end, a scientific technique that can combine the multi-dimensional safety performance indicators (SPIs) into an overall index, and subsequently can identify the 'best-in-class' is urgently required. In this paper, the Entropy-embedded RSR (Rank-sum ratio), an innovative, scientific and systematic methodology is investigated with the aim of conducting the above two core tasks in an integrative and concise procedure, more specifically in a 'one-stop' way. Using a combination of results from other methods (e.g. the SUNflower approach) and other measures (e.g. Human Development Index) as a relevant reference, a given set of European countries are robustly ranked and grouped into several classes based on the composite Road Safety Index. Within each class the 'best-in-class' is then identified. By benchmarking road safety performance, the results serve to promote best practice, encourage the adoption of successful road safety strategies and measures and, more importantly, inspire the kind of political leadership needed to create a road transport system that maximizes safety. Copyright © 2015 Elsevier Ltd. All rights reserved.

Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials: The OMERACT Safety Working Group.

PubMed

Klokker, Louise; Tugwell, Peter; Furst, Daniel E; Devoe, Dan; Williamson, Paula; Terwee, Caroline B; Suarez-Almazor, Maria E; Strand, Vibeke; Woodworth, Thasia; Leong, Amye L; Goel, Niti; Boers, Maarten; Brooks, Peter M; Simon, Lee S; Christensen, Robin

2017-12-01

Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. The safety issue has previously been discussed at OMERACT, but without a consistent approach to ensure harms were included in COS. Our methods include (1) identifying harmful outcomes in trials of interventions studied in patients with rheumatic diseases by a systematic literature review, (2) identifying components of safety that should be measured in such trials by use of a patient-driven approach including qualitative data collection and statistical organization of data, and (3) developing a COS through consensus processes including everyone involved. Members of OMERACT including patients, clinicians, researchers, methodologists, and industry representatives reached consensus on the need to continue the efforts on developing a COS for safety in rheumatology trials. There was a general agreement about the need to identify safety-related outcomes that are meaningful to patients, framed in terms that patients consider relevant so that they will be able to make informed decisions. The OMERACT Safety Working Group will advance the work previously done within OMERACT using a new patient-driven approach.

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.

Safety pharmacology — Current and emerging concepts

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

Hamdam, Junnat; Sethu, Swaminathan; Smith, Trevor

2013-12-01

Safety pharmacology (SP) is an essential part of the drug development process that aims to identify and predict adverse effects prior to clinical trials. SP studies are described in the International Conference on Harmonisation (ICH) S7A and S7B guidelines. The core battery and supplemental SP studies evaluate effects of a new chemical entity (NCE) at both anticipated therapeutic and supra-therapeutic exposures on major organ systems, including cardiovascular, central nervous, respiratory, renal and gastrointestinal. This review outlines the current practices and emerging concepts in SP studies including frontloading, parallel assessment of core battery studies, use of non-standard species, biomarkers, and combiningmore » toxicology and SP assessments. Integration of the newer approaches to routine SP studies may significantly enhance the scope of SP by refining and providing mechanistic insight to potential adverse effects associated with test compounds. - Highlights: • SP — mandatory non-clinical risk assessments performed during drug development. • SP organ system studies ensure the safety of clinical participants in FiH trials. • Frontloading in SP facilitates lead candidate drug selection. • Emerging trends: integrating SP-Toxicological endpoints; combined core battery tests.« less

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

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

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

1999-09-27

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

Safety First: Feather, Feet, and Fin Safety in the Classroom

ERIC Educational Resources Information Center

Roy, Ken

2014-01-01

Pet birds, fish, reptiles, and mammals--all are often found in elementary classrooms because of the wide variety of opportunities they provide for exciting teaching and learning experiences. Applications of the opportunities these organisms can provide is reflected in the "NGSS" Life Science progression of disciplinary core ideas,…

Highway Maintenance Equipment Operator: Basic Core. Training Materials.

ERIC Educational Resources Information Center

Perky, Sandra Dutreau; And Others

This basic core curriculum is part of a three-part series of instructional guides designed for use in teaching a course in highway maintenance equipment operation. Addressed in the individual units of the curriculum, after an orientation unit, are safety; basic math; basic hand tools; procedures for loading. lashing, and unloading equipment;…

High Resolution Continuous Flow Analysis System for Polar Ice Cores

NASA Astrophysics Data System (ADS)

Dallmayr, Remi; Azuma, Kumiko; Yamada, Hironobu; Kjær, Helle Astrid; Vallelonga, Paul; Azuma, Nobuhiko; Takata, Morimasa

2014-05-01

In the last decades, Continuous Flow Analysis (CFA) technology for ice core analyses has been developed to reconstruct the past changes of the climate system 1), 2). Compared with traditional analyses of discrete samples, a CFA system offers much faster and higher depth resolution analyses. It also generates a decontaminated sample stream without time-consuming sample processing procedure by using the inner area of an ice-core sample.. The CFA system that we have been developing is currently able to continuously measure stable water isotopes 3) and electrolytic conductivity, as well as to collect discrete samples for the both inner and outer areas with variable depth resolutions. Chemistry analyses4) and methane-gas analysis 5) are planned to be added using the continuous water stream system 5). In order to optimize the resolution of the current system with minimal sample volumes necessary for different analyses, our CFA system typically melts an ice core at 1.6 cm/min. Instead of using a wire position encoder with typical 1mm positioning resolution 6), we decided to use a high-accuracy CCD Laser displacement sensor (LKG-G505, Keyence). At the 1.6 cm/min melt rate, the positioning resolution was increased to 0.27mm. Also, the mixing volume that occurs in our open split debubbler is regulated using its weight. The overflow pumping rate is smoothly PID controlled to maintain the weight as low as possible, while keeping a safety buffer of water to avoid air bubbles downstream. To evaluate the system's depth-resolution, we will present the preliminary data of electrolytic conductivity obtained by melting 12 bags of the North Greenland Eemian Ice Drilling (NEEM) ice core. The samples correspond to different climate intervals (Greenland Stadial 21, 22, Greenland Stadial 5, Greenland Interstadial 5, Greenland Interstadial 7, Greenland Stadial 8). We will present results for the Greenland Stadial -8, whose depths and ages are between 1723.7 and 1724.8 meters, and 35.520 to 35.636 kyr b2k 7), respectively. The results show the conductivity measured upstream and downstream of the debubbler. We will calculate the depth resolution of our system and compare it with earlier studies. 1) Bigler at al, "Optimization of High-Resolution Continuous Flow Analysis For Transient Climate Signals in Ice Cores". Environ. Sci. Technol. 2011, 45, 4483-4489 2) Kaufmann et al, "An Improved Continuous Flow Analysis System for High Resolution Field Measurements on Ice Cores". Environmental Environ. Sci. Technol. 2008, 42, 8044-8050 3) Gkinis, V., T. J. Popp, S. J. Johnsen and T, Blunier, 2010: A continuous stream flash evaporator for the calibration of an IR cavity ring down spectrometer for the isotopic analysis of water. Isotopes in Environmental and Health Studies, 46(4), 463-475. 4) McConnell et al, "Continuous ice-core chemical analyses using inductively coupled plasma mass spectrometry. Environ. Sci. Technol. 2002, 36, 7-11 5) Rhodes et al, "Continuous methane measurements from a late Holocene Greenland ice core : Atmospheric and in-situ signals" Earth and Planetary Science Letters. 2013, 368, 9-19 6) Breton et al, "Quantifying Signal Dispersion in a Hybrid Ice Core Melting System". Environ. Sci. Technol. 2012, 46, 11922-11928 7) Rasmussen et al, " A first chronology for the NEEM ice core". Climate of the Past. 2013, 9, 2967--3013

Core characterization of the new CABRI Water Loop Facility

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

Ritter, G.; Rodiac, F.; Beretz, D.

2011-07-01

The CABRI experimental reactor is located at the Cadarache nuclear research center, southern France. It is operated by the Atomic Energy Commission (CEA) and devoted to IRSN (Institut de Radioprotection et de Surete Nucleaire) safety programmes. It has been successfully operated during the last 30 years, enlightening the knowledge of FBR and LWR fuel behaviour during Reactivity Insertion Accident (RIA) and Loss Of Coolant Accident (LOCA) transients in the frame of IPSN (Institut de Protection et de Surete Nucleaire) and now IRSN programmes devoted to reactor safety. This operation was interrupted in 2003 to allow for a whole facility renewalmore » programme for the need of the CABRI International Programme (CIP) carried out by IRSN under the OECD umbrella. The principle of operation of the facility is based on the control of {sup 3}He, a major gaseous neutron absorber, in the core geometry. The purpose of this paper is to illustrate how several dosimetric devices have been set up to better characterize the core during the upcoming commissioning campaign. It presents the schemes and tools dedicated to core characterization. (authors)« less

A patient safety objective structured clinical examination.

PubMed

Singh, Ranjit; Singh, Ashok; Fish, Reva; McLean, Don; Anderson, Diana R; Singh, Gurdev

2009-06-01

There are international calls for improving education for health care workers around certain core competencies, of which patient safety and quality are integral and transcendent parts. Although relevant teaching programs have been developed, little is known about how best to assess their effectiveness. The objective of this work was to develop and implement an objective structured clinical examination (OSCE) to evaluate the impact of a patient safety curriculum. The curriculum was implemented in a family medicine residency program with 47 trainees. Two years after commencing the curriculum, a patient safety OSCE was developed and administered at this program and, for comparison purposes, to incoming residents at the same program and to residents at a neighboring residency program. All 47 residents exposed to the training, all 16 incoming residents, and 10 of 12 residents at the neighboring program participated in the OSCE. In a standardized patient case, error detection and error disclosure skills were better among trained residents. In a chart-based case, trained residents showed better performance in identifying deficiencies in care and described more appropriate means of addressing them. Third year residents exposed to a "Systems Approach" course performed better at system analysis and identifying system-based solutions after the course than before. Results suggest increased systems thinking and inculcation of a culture of safety among residents exposed to a patient safety curriculum. The main weaknesses of the study are its small size and suboptimal design. Much further investigation is needed into the effectiveness of patient safety curricula.

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

Applications of the RELAP5 code to the station blackout transients at the Browns Ferry Unit One Plant

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

Schultz, R.R.; Wagoner, S.R.

1983-01-01

As a part of the charter of the Severe Accident Sequence Analysis (SASA) Program, station blackout transients have been analyzed using a RELAP5 model of the Browns Ferry Unit 1 Plant. The task was conducted as a partial fulfillment of the needs of the US Nuclear Regulatory Commission in examining the Unresolved Safety Issue A-44: Station Blackout (1) the station blackout transients were examined (a) to define the equipment needed to maintain a well cooled core, (b) to determine when core uncovery would occur given equipment failure, and (c) to characterize the behavior of the vessel thermal-hydraulics during the stationmore » blackout transients (in part as the plant operator would see it). These items are discussed in the paper. Conclusions and observations specific to the station blackout are presented.« less

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

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

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

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

Gillespie, B.M.; Stromatt, R.W.; Ross, G.A.

This data package contains the results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization of samples for the 101-SY Hydrogen Safety Project. The samples were submitted for analysis by Westinghouse Hanford Company (WHC) under the Technical Project Plan (TPP) 17667 and the Quality Assurance Plan MCS-027. They came from a core taken during Window C'' after the May 1991 gas release event. The analytical procedures required for analysis were defined in the Test Instructions (TI) prepared by the PNL 101-SY Analytical Chemistry Laboratory (ACL) Project Management Office in accordance with the TPP and the QA Plan. The requestedmore » analysis for these samples was volatile organic analysis. The quality control (QC) requirements for each sample are defined in the Test Instructions for each sample. The QC requirements outlined in the procedures and requested in the WHC statement of work were followed.« less

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

Gillespie, B.M.; Stromatt, R.W.; Ross, G.A.

This data package contains the results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization of samples for the 101-SY Hydrogen Safety Project. The samples were submitted for analysis by Westinghouse Hanford Company (WHC) under the Technical Project Plan (TPP) 17667 and the Quality Assurance Plan MCS-027. They came from a core taken during Window ``C`` after the May 1991 gas release event. The analytical procedures required for analysis were defined in the Test Instructions (TI) prepared by the PNL 101-SY Analytical Chemistry Laboratory (ACL) Project Management Office in accordance with the TPP and the QA Plan. The requestedmore » analysis for these samples was volatile organic analysis. The quality control (QC) requirements for each sample are defined in the Test Instructions for each sample. The QC requirements outlined in the procedures and requested in the WHC statement of work were followed.« less

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

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

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

Bricklaying Curriculum: Basic Core. Instructional Materials. Revised.

ERIC Educational Resources Information Center

Turcotte, Raymond J.; Hendrix, Laborn J.

This volume, the first in a two-volume core curriculum, is designed for use in teaching a course in basic bricklaying. Included in the introductory section of the guide are units on the free enterprise system, the economics of free enterprise, industry orientation, ways of becoming a good leader, job advancement, and safety and first aid. The next…

Agricultural Mechanics Unit for Plant Science Core Curriculum. Volume 15, Number 4. Instructor's Guide.

ERIC Educational Resources Information Center

Linhardt, Richard E.; Hunter, Bill

This instructor's guide is intended for use in teaching the agricultural mechanics unit of a plant science core curriculum. Covered in the individual units of the guide are the following topics: arc welding (following safety procedures, controlling distortion, selecting and caring for electrodes, identifying the material to be welded, and welding…

A CORE PROBLEM ON SAFETY, TEACHING CORE.

ERIC Educational Resources Information Center

LONG, SADIE

A TRAFFIC ACCIDENT MOTIVATED A GROUP OF EIGHTH-GRADE CIVICS STUDENTS TO FORM DIFFERENT COMMITTEES TO OBTAIN A TRAFFIC GUARD FOR THEIR SCHOOL. COMMITTEES WERE FORMED TO CONTACT THE POLICE DEPARTMENT BY LETTER AND INTERVIEW, TO TAKE PICTURES OF THE STREET CORNER AND MAKE CHARTS SHOWING ITS DANGER, TO DO RESEARCH ON CONSTITUTIONAL RIGHTS OF CITIZENS,…

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Preparing for the Update of Vermont’s Strategic Highway Safety Plan : Proceedings from the Federal Highway Administration’s Peer-to-Peer Exchange Program

DOT National Transportation Integrated Search

2011-01-01

This report provides a summary of a peer exchange sponsored by the Vermont Agency of Transportation (VTrans). The peer exchange convened Vermonts Strategic Highway Safety Plan (SHSP) Core Group to discuss the strengths and weaknesses of Vermont...

29 CFR 1926.104 - Safety belts, lifelines, and lanyards.

Code of Federal Regulations, 2014 CFR

2014-07-01

... be subjected to cutting or abrasion, shall be a minimum of 7/8-inch wire core manila rope. For all other lifeline applications, a minimum of 3/4-inch manila or equivalent, with a minimum breaking... nominal breaking strength of 5,400 pounds. (e) All safety belt and lanyard hardware shall be drop forged...

29 CFR 1926.104 - Safety belts, lifelines, and lanyards.

Code of Federal Regulations, 2012 CFR

2012-07-01

... be subjected to cutting or abrasion, shall be a minimum of 7/8-inch wire core manila rope. For all other lifeline applications, a minimum of 3/4-inch manila or equivalent, with a minimum breaking... nominal breaking strength of 5,400 pounds. (e) All safety belt and lanyard hardware shall be drop forged...

29 CFR 1926.104 - Safety belts, lifelines, and lanyards.

Code of Federal Regulations, 2013 CFR

2013-07-01

... be subjected to cutting or abrasion, shall be a minimum of 7/8-inch wire core manila rope. For all other lifeline applications, a minimum of 3/4-inch manila or equivalent, with a minimum breaking... nominal breaking strength of 5,400 pounds. (e) All safety belt and lanyard hardware shall be drop forged...

Establishing a Safe School Culture: An Examination of Current Practices in K through 12 Leadership

ERIC Educational Resources Information Center

Kelly, Zanita V.

2016-01-01

School improvement plans and major reform initiatives most often target core academic competencies. They might include strategies to improve the physical safety of school campuses, but they rarely include discussions about creating psychologically safe environments. School safety has garnered national attention in the aftermath of violent high…

Initial development of a practical safety audit tool to assess fleet safety management practices.

PubMed

Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

2012-07-01

Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

DEVICE FOR CONTROLLING INSERTION OF ROD

DOEpatents

Beaty, B.J.

1958-10-14

A device for rapidly inserting a safety rod into a nuclear reactor upon a given signal or in the event of a power failure in order to prevent the possibility of extensive damage caused by a power excursion is described. A piston is slidably mounted within a vertical cylinder with provision for an electromagnetic latch at the top of the cylinder. This assembly, with a safety rod attached to the piston, is mounted over an access port to the core region of the reactor. The piston is normally latched at the top of the cylinder with the safety rod clear of the core area, however, when the latch is released, the piston and rod drop by their own weight to insert the rod. Vents along the side of the cylinder permit the escape of the air entrapped under the piston over the greater part of the distance, however, at the end of the fall the entrapped air is compressed thereby bringing the safety rod gently to rest, thus providing for a rapid automatic insertion of the rod with a minimum of structural shock.

DairyBeef: maximizing quality and profits--a consistent food safety message.

PubMed

Moore, D A; Kirk, J H; Klingborg, D J; Garry, F; Wailes, W; Dalton, J; Busboom, J; Sams, R W; Poe, M; Payne, M; Marchello, J; Looper, M; Falk, D; Wright, T

2004-01-01

To respond to meat safety and quality issues in dairy market cattle, a collaborative project team for 7 western states was established to develop educational resources providing a consistent meat safety and quality message to dairy producers, farm advisors, and veterinarians. The team produced an educational website and CD-ROM course that included videos, narrated slide sets, and on-farm tools. The objectives of this course were: 1) to help producers and their advisors understand market cattle food safety and quality issues, 2) help maintain markets for these cows, and 3) help producers identify ways to improve the quality of dairy cattle going to slaughter. DairyBeef. Maximizing Quality & Profits consists of 6 sections, including 4 core segments. Successful completion of quizzes following each core segment is required for participants to receive a certificate of completion. A formative evaluation of the program revealed the necessity for minor content and technological changes with the web-based course. All evaluators considered the materials relevant to dairy producers. After editing, course availability was enabled in February, 2003. Between February and May, 2003, 21 individuals received certificates of completion.

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

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

The non-resonant kink modes triggering strong sawtooth-like crashes in the EAST tokamak

NASA Astrophysics Data System (ADS)

Li, Erzhong; Igochine, V.; Dumbrajs, O.; Xu, L.; Chen, K.; Shi, T.; Hu, L.

2014-12-01

Evolution of the safety factor (q) profile during L-H transitions in the Experimental Advanced Superconducting Tokamak (EAST) was accompanied by strong core crashes prior to regular sawtooth behavior. These crashes appeared in the absence of q = 1 (q is the safety factor) rational surface inside the plasma. Analysis indicates that the m/n = 2/1 tearing mode is destabilized and phase-locked with the m/n = 1/1 non-resonant kink mode (the q = 1 rational surface is absent) due to the self-consistent evolution of plasma profiles as the L-H transition occurs (m and n are the poloidal and toroidal mode numbers, respectively). The growing m/n = 1/1 mode destabilizes the m/n = 2/2 kink mode which eventually triggers the strong crash due to an anomalous heat conductivity, as predicted by the transport model of stochastic magnetic fields using experimental parameters. It is also shown that the magnetic topology changes with the amplitude of m/n = 2/2 mode and the value of center safety factor in a reasonable range.

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

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

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

Cascade Distillation System Design for Safety and Mission Assurance

NASA Technical Reports Server (NTRS)

Sargusingh, Miriam J.; Callahan, Michael R.

2015-01-01

Per the NASA Human Health, Life Support and Habitation System Technology Area 06 report "crewed missions venturing beyond Low-Earth Orbit (LEO) will require technologies with improved reliability, reduced mass, self-sufficiency, and minimal logistical needs as an emergency or quick-return option will not be feasible." To meet this need, the development team of the second generation Cascade Distillation System (CDS 2.0) opted a development approach that explicitely incorporate consideration of safety, mission assurance, and autonomy. The CDS 2.0 prelimnary design focused on establishing a functional baseline that meets the CDS core capabilities and performance. The critical design phase is now focused on incorporating features through a deliberative process of establishing the systems failure modes and effects, identifying mitigative strategies, and evaluating the merit of the proposed actions through analysis and test. This paper details results of this effort on the CDS 2.0 design.

Cascade Distillation System Design for Safety and Mission Assurance

NASA Technical Reports Server (NTRS)

Sarguisingh, Miriam; Callahan, Michael R.; Okon, Shira

2015-01-01

Per the NASA Human Health, Life Support and Habitation System Technology Area 06 report "crewed missions venturing beyond Low-Earth Orbit (LEO) will require technologies with improved reliability, reduced mass, self-sufficiency, and minimal logistical needs as an emergency or quick-return option will not be feasible".1 To meet this need, the development team of the second generation Cascade Distillation System (CDS 2.0) chose a development approach that explicitly incorporate consideration of safety, mission assurance, and autonomy. The CDS 2.0 preliminary design focused on establishing a functional baseline that meets the CDS core capabilities and performance. The critical design phase is now focused on incorporating features through a deliberative process of establishing the systems failure modes and effects, identifying mitigation strategies, and evaluating the merit of the proposed actions through analysis and test. This paper details results of this effort on the CDS 2.0 design.

A Statistics-Based Cracking Criterion of Resin-Bonded Silica Sand for Casting Process Simulation

NASA Astrophysics Data System (ADS)

Wang, Huimin; Lu, Yan; Ripplinger, Keith; Detwiler, Duane; Luo, Alan A.

2017-02-01

Cracking of sand molds/cores can result in many casting defects such as veining. A robust cracking criterion is needed in casting process simulation for predicting/controlling such defects. A cracking probability map, relating to fracture stress and effective volume, was proposed for resin-bonded silica sand based on Weibull statistics. Three-point bending test results of sand samples were used to generate the cracking map and set up a safety line for cracking criterion. Tensile test results confirmed the accuracy of the safety line for cracking prediction. A laboratory casting experiment was designed and carried out to predict cracking of a cup mold during aluminum casting. The stress-strain behavior and the effective volume of the cup molds were calculated using a finite element analysis code ProCAST®. Furthermore, an energy dispersive spectroscopy fractographic examination of the sand samples confirmed the binder cracking in resin-bonded silica sand.

Complexity of the international agro-food trade network and its impact on food safety.

PubMed

Ercsey-Ravasz, Mária; Toroczkai, Zoltán; Lakner, Zoltán; Baranyi, József

2012-01-01

With the world's population now in excess of 7 billion, it is vital to ensure the chemical and microbiological safety of our food, while maintaining the sustainability of its production, distribution and trade. Using UN databases, here we show that the international agro-food trade network (IFTN), with nodes and edges representing countries and import-export fluxes, respectively, has evolved into a highly heterogeneous, complex supply-chain network. Seven countries form the core of the IFTN, with high values of betweenness centrality and each trading with over 77% of all the countries in the world. Graph theoretical analysis and a dynamic food flux model show that the IFTN provides a vehicle suitable for the fast distribution of potential contaminants but unsuitable for tracing their origin. In particular, we show that high values of node betweenness and vulnerability correlate well with recorded large food poisoning outbreaks.

An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

PubMed

Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

2017-05-01

Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

FINAL SAFETY ANALYSIS REPORT--SNAP 1A RADIOISOTOPE FUELED THERMOELECTRIC GENERATOR

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

Dix, G.P.

1960-06-30

The safety aspects involved in utilizing the Task 2 radioisotope-powered thermoelectric generator in a terrestrial satellite are described. It is based upon a generalized satellite mission having a 600-day orbital lifetime. A description of the basic design of the generator is presented in order to establish the analytical model. This includes the generator design, radiocerium fuel properties, and the fuel core. The transport of the generator to the launch site is examined, including the shipping cask, shipping procedures, and shipping hazards. A description of ground handling and vehicle integration is presented including preparation for fuel transfer, transfer, mating of generatorsmore » to final stage, mating final stage to booster, and auxiliary support equipment. The flight vehicle is presented to complete the analytical model. Contained in this chapter are descriptions of the booster-sustainer, final stage, propellants, and built-in safety systems. The typical missile range is examined with respect to the launch complex and range safety characteristics. The shielding of the fuel is discussed and includes both dose rates and shield thicknesses required. The bare core, shielded generator, fuel transfer operation and dose rates for accidental conditions are treated. mechanism of re-entry from the successful mission is covered. Radiocerium inventories with respect to time and the chronology of re-entry are specifically treated. The multiplicity of conditions for aborted missions is set forth. The definition of aborted missions is treated first in order to present the initial conditions. Following this, a definition of the forces imposed upon the generator is presented. The aborted missions is presented. A large number of initial vehicle failure cases is narrowed down into categories of consequences. Since stratospheric injection of fuel results in cases where the fuel is not contained after re-entry, an extensive discussion of the fall-out mechanism is presented. (auth)« less

Narcotics reduction, quality and safety in gynecologic oncology surgery in the first year of enhanced recovery after surgery protocol implementation.

PubMed

Bergstrom, Jennifer E; Scott, Marla E; Alimi, Yewande; Yen, Ting-Tai; Hobson, Deborah; Machado, Karime K; Tanner, Edward J; Fader, Amanda N; Temkin, Sarah M; Wethington, Stephanie; Levinson, Kimberly; Sokolinsky, Sam; Lau, Brandyn; Stone, Rebecca L

2018-06-01

Enhanced Recovery After Surgery (ERAS) programs are mechanisms for achieving value-based improvements in surgery. This report provides a detailed analysis of the impact of an ERAS program on patient outcomes as well as quality and safety measures during implementation on a gynecologic oncology service at a major academic medical center. A retrospective review of gynecologic oncology patients undergoing elective laparotomy during the implementation phase of an ERAS program (January 2016 through December 2016) was performed. Patient demographics, surgical variables, postoperative outcomes, and adherence to core safety measures, including antimicrobial and venous thromboembolism (VTE) prophylaxis, were compared to a historical patient cohort (January 2015 through December 2015). Statistical analyses were performed using t-tests, Wilcoxon rank sum tests, and Chi squared tests. The inaugural 109 ERAS program participants were compared to a historical patient cohort (n=158). There was no difference in BMI, race, malignancy, or complexity of procedure between cohorts. ERAS patients required less narcotics (70.7 vs 127.4, p=0.007, oral morphine equivalents) and PCA use (32.1% vs. 50.6%, p=0.002). Despite this substantial reduction in narcotics, ERAS patients did not report more pain and in fact reported significantly less pain by postoperative day 3. There were no differences in length of stay (5days), complication rates (13.8% vs. 20.3%, p=0.17) or 30-day readmission rates (9.5 vs 11.9%, p=0.54) between ERAS and historical patients, respectively. Compliance with antimicrobial prophylaxis was 97.2%. However, 33.9% of ERAS patients received substandard preoperative VTE prophylaxis. ERAS program implementation resulted in reductions in narcotic requirements and PCA use without changes in length of stay or readmission rates. Compliance should be diligently audited during the implementation phase of ERAS programs, with special attention to adherence to pre-existing core safety measures. Copyright © 2018 Elsevier Inc. All rights reserved.

Eight-year follow-up data from the U.S. clinical trial for Sientra's FDA-approved round and shaped implants with high-strength cohesive silicone gel.

PubMed

Stevens, W Grant; Harrington, Jennifer; Alizadeh, Kaveh; Broadway, David; Zeidler, Kamakshi; Godinez, Tess B

2015-05-01

On March 9, 2012, the Food and Drug Administration (FDA) approved Sientra's premarket approval application for its portfolio of silicone gel breast implants based on their review of Sientra's 3-year study data from the largest pivotal silicone gel breast implant study to date. This included the first approval of shaped breast implants in the United States. The authors provide an update to the 8-year safety and effectiveness of the Sientra High-Strength silicone gel breast implants. The Sientra Core study is an ongoing 10 year open-label, prospective, multi-center clinical study, which includes 1788 patients implanted with 3506 Sientra implants across four indications (Primary Augmentation, Revision Augmentation, Primary Reconstruction, and Revision Reconstruction). For the safety analysis, the incidence of post-operative complications, including all breast implant-related adverse effects (eg, infection, asymmetry), was estimated based on Kaplan-Meier risk rates. The effectiveness analyses include surgeon and patient satisfaction and changes in bra/cup size. Through 8 years, the overall risk of rupture was 4.6%, the risk of capsular contracture was 11.8% (rates were lower when using True Texture™), and the risk of reoperation was 28.3%. Out of the 580 reoperations in 456 patients, over half of all reoperations were due to cosmetic reasons (n = 299). The most common reasons for reoperation were capsular contracture (19.0%), style and/or size change (18.4%), and asymmetry (8.8%). Patient satisfaction remains high through 8 years, with 87% indicating that their breast implants make them feel more feminine than prior to enrollment. Safety data from the FDA Core study continues to support a comprehensive safety and effectiveness profile of Sientra's portfolio of round and shaped implants through 8 years. 3 Therapeutic. © 2015 The American Society for Aesthetic Plastic Surgery, Inc. Reprints and permission: journals.permissions@oup.com.

Creativity and connections: the future of nursing education and practice: the Massachusetts Initiative.

PubMed

Sroczynski, Maureen; Gravlin, Gayle; Route, Paulette Seymour; Hoffart, Nancy; Creelman, Patricia

2011-01-01

Education and practice partnerships are key to effective academic program design and implementation in a time of decreasing supply and increasing demands on the nursing profession. An integrated education/practice competency model can positively impact patient safety, improve patient care, increase retention, and ensure a sufficient and competent nursing workforce, which is paramount to survival of the health care system. Through the contributions of nursing leaders from the broad spectrum of nursing and industry organizations within the state, the Massachusetts Nurse of the Future project developed a competency-based framework for the future design of nursing educational programs to meet current and future practice needs. The Massachusetts Nurse of the Future Nursing Core Competencies(©) expand on the Institute of Medicine's core competencies for all health care professionals and the Quality and Safety Education for Nurses competencies for quality and safety to define the expectations for all professional nurses of the future. The Massachusetts Nurse of the Future Nursing Core Competencies define the knowledge, attitude, and skills required as the minimal expectations for initial nursing practice following completion of a prelicensure professional nursing education program. These competencies are now being integrated into new models for seamless, coordinated nursing curriculum and transition into practice within the state and beyond. Copyright © 2011 Elsevier Inc. All rights reserved.

75 FR 53985 - Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit 3; Temporary...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-02

... are authorized by law, will not present an undue risk to public health or safety, and are consistent... Public Health and Safety The underlying purpose of 10 CFR 50.46 is to establish acceptance criteria for... (LOCA) and non-LOCA criteria, mechanical design, thermal hydraulics, seismic, core physics, and...

29 CFR 1960.12 - Dissemination of occupational safety and health program information.

Code of Federal Regulations, 2011 CFR

2011-07-01

... establishment, and keep posted, a poster informing employees of the provisions of the Act, Executive Order 12196... furnish the core text of a poster to agencies. Each agency shall add the following items: (1) Details of...) Relevant information about any agency safety and health committees. Such posters and additions shall not be...

29 CFR 1960.12 - Dissemination of occupational safety and health program information.

Code of Federal Regulations, 2013 CFR

2013-07-01

... establishment, and keep posted, a poster informing employees of the provisions of the Act, Executive Order 12196... furnish the core text of a poster to agencies. Each agency shall add the following items: (1) Details of...) Relevant information about any agency safety and health committees. Such posters and additions shall not be...

29 CFR 1960.12 - Dissemination of occupational safety and health program information.

Code of Federal Regulations, 2012 CFR

2012-07-01

... establishment, and keep posted, a poster informing employees of the provisions of the Act, Executive Order 12196... furnish the core text of a poster to agencies. Each agency shall add the following items: (1) Details of...) Relevant information about any agency safety and health committees. Such posters and additions shall not be...

29 CFR 1960.12 - Dissemination of occupational safety and health program information.

Code of Federal Regulations, 2010 CFR

2010-07-01

... establishment, and keep posted, a poster informing employees of the provisions of the Act, Executive Order 12196... furnish the core text of a poster to agencies. Each agency shall add the following items: (1) Details of...) Relevant information about any agency safety and health committees. Such posters and additions shall not be...

29 CFR 1960.12 - Dissemination of occupational safety and health program information.

Code of Federal Regulations, 2014 CFR

2014-07-01

... establishment, and keep posted, a poster informing employees of the provisions of the Act, Executive Order 12196... furnish the core text of a poster to agencies. Each agency shall add the following items: (1) Details of...) Relevant information about any agency safety and health committees. Such posters and additions shall not be...

Study on Safety Monitoring System for Submarine Power Cable on the Basis of AIS and Radar Technology

NASA Astrophysics Data System (ADS)

Jie, Wang; Yao-Tian, Fan

Through analyzing the risks of submarine power cable, the highest risk to damage the cable identified is from ship. Based on concept of Vessel Traffic Management Information Systems, the three core sub-systems of safety monitoring system for submarine power cable were studied and described, also some suggestions were given.

10 CFR Appendix E to Part 50 - Emergency Planning and Preparedness for Production and Utilization Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... could communicate with a safety system. In this case, appropriate isolation devices would be required at..., feedwater flow, and reactor power; (2) Safety injection: Reactor core isolation cooling flow, high-pressure... data points identified in the ERDS Data Point Library 9 (site specific data base residing on the ERDS...

10 CFR Appendix E to Part 50 - Emergency Planning and Preparedness for Production and Utilization Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... could communicate with a safety system. In this case, appropriate isolation devices would be required at..., feedwater flow, and reactor power; (2) Safety injection: Reactor core isolation cooling flow, high-pressure... data points identified in the ERDS Data Point Library 9 (site specific data base residing on the ERDS...

The Perception of Safety between Drinkers and Non-Drinkers among U.S. College Students

ERIC Educational Resources Information Center

Walter, Gayle; Florkowski, David; Anderson, Peter; Dunn, Micheal

2014-01-01

Increasing episodes of campus violence have warranted an investigation into college students' perception of safety on campus. In this study, 56,811 students responded to the Core Alcohol and Drug Survey during the 2010 academic school year. Numerous universities administered the survey and students completed the survey either in class or…

Using resources for scientific-driven pharmacovigilance: from many product safety documents to one product safety master file.

PubMed

Furlan, Giovanni

2012-08-01

Current regulations require a description of the overall safety profile or the specific risks of a drug in multiple documents such as the Periodic and Development Safety Update Reports, Risk Management Plans (RMPs) and Signal Detection Reports. In a resource-constrained world, the need for preparing multiple documents reporting the same information results in shifting the focus from a thorough scientific and medical evaluation of the available data to maintaining compliance with regulatory timelines. Since the aim of drug safety is to understand and characterize product issues to take adequate risk minimization measures rather than to comply with bureaucratic requirements, there is the need to avoid redundancy. In order to identify core drug safety activities that need to be undertaken to protect patient safety and reduce the number of documents reporting the results of these activities, the author has reviewed the main topics included in the drug safety guidelines and templates. The topics and sources that need to be taken into account in the main regulatory documents have been found to greatly overlap and, in the future, as a result of the new Periodic Safety Update Report structure and requirements, in the author's opinion this overlap is likely to further increase. Many of the identified inter-document differences seemed to be substantially formal. The Development Safety Update Report, for example, requires separate presentation of the safety issues emerging from different sources followed by an overall evaluation of each safety issue. The RMP, instead, requires a detailed description of the safety issues without separate presentation of the evidence derived from each source. To some extent, however, the individual documents require an in-depth analysis of different aspects; the RMP, for example, requires an epidemiological description of the indication for which the drug is used and its risks. At the time of writing this article, this is not specifically required by other documents. The author has identified signal detection (intended not only as adverse event disproportionate reporting, but including non-clinical, laboratory, clinical analysis data and literature screening) and characterization as the basis for the preparation of all drug safety documents, which can be viewed as different ways of presenting the results of this activity. Therefore, the author proposes to merge all the aggregate reports required by current regulations into a single document - the Drug Safety Master File. This report should contain all the available information, from any source, regarding the potential and identified risks of a drug. It should be a living document updated and submitted to regulatory authorities on an ongoing basis.

Bayesian network representing system dynamics in risk analysis of nuclear systems

NASA Astrophysics Data System (ADS)

Varuttamaseni, Athi

2011-12-01

A dynamic Bayesian network (DBN) model is used in conjunction with the alternating conditional expectation (ACE) regression method to analyze the risk associated with the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed operation in the Zion-1 nuclear power plant. The use of the DBN allows the joint probability distribution to be factorized, enabling the analysis to be done on many simpler network structures rather than on one complicated structure. The construction of the DBN model assumes conditional independence relations among certain key reactor parameters. The choice of parameter to model is based on considerations of the macroscopic balance statements governing the behavior of the reactor under a quasi-static assumption. The DBN is used to relate the peak clad temperature to a set of independent variables that are known to be important in determining the success of the feed and bleed operation. A simple linear relationship is then used to relate the clad temperature to the core damage probability. To obtain a quantitative relationship among different nodes in the DBN, surrogates of the RELAP5 reactor transient analysis code are used. These surrogates are generated by applying the ACE algorithm to output data obtained from about 50 RELAP5 cases covering a wide range of the selected independent variables. These surrogates allow important safety parameters such as the fuel clad temperature to be expressed as a function of key reactor parameters such as the coolant temperature and pressure together with important independent variables such as the scram delay time. The time-dependent core damage probability is calculated by sampling the independent variables from their probability distributions and propagate the information up through the Bayesian network to give the clad temperature. With the knowledge of the clad temperature and the assumption that the core damage probability has a one-to-one relationship to it, we have calculated the core damage probably as a function of transient time. The use of the DBN model in combination with ACE allows risk analysis to be performed with much less effort than if the analysis were done using the standard techniques.

Lot-to-lot consistency study of the fully liquid pentavalent DTwP-HepB-Hib vaccine Quinvaxem® demonstrating clinical equivalence, suitability of the vaccine as a booster and concomitant administration with measles vaccine

PubMed Central

Aspinall, Sanet; Traynor, Deirdre; Bedford, Philip; Hartmann, Katharina

2012-01-01

This double-blind, randomized study evaluated the immunogenicity and safety of three production lots of the fully liquid combination DTwP-Hep-Hib vaccine, Quinvaxem® (Crucell, The Netherlands) in 360 healthy infants aged 42–64 d old given at 6, 10 and 14 weeks of age (Core Study). The Core Study was followed by an open-label Booster Phase evaluating immunogenicity and safety of a booster dose of Quinvaxem® given with either concomitant or deferred measles vaccine in 227 infants who completed the Core Study. One month after the third dose of Quinvaxem® immune responses reflecting seroprotection or seroconversion were observed in more than 90% of infants for all three vaccine lots. Quinvaxem® elicited a strong booster response as demonstrated by a large increase in antibodies against all antigens, which appeared to be unaffected by concomitant administration of the measles vaccine. Safety results were in line with previous reports for Quinvaxem® with no unexpected adverse events (AEs) being reported. In the Core Study and Booster Phase, Quinvaxem® was well tolerated. No study vaccine-related serious AEs were reported. Thus, Quinvaxem® was immunogenic and well-tolerated when administered to infants according to a 6–10–14 week vaccination schedule. The three production lots had consistent reactogenicity and immunogenicity profiles. The booster dose of Quinvaxem® was also immunogenic and safe, regardless of whether a monovalent measles vaccine was administered concomitantly or one month later. PMID:22854660

A Novel In Vivo Protocol for Molecular Study of Radiation-Induced Fibrosis in Head and Neck Cancer Patients.

PubMed

Krisciunas, Gintas P; Platt, Michael; Trojanowska, Maria; Grillone, Gregory A; Haines, Paul C; Langmore, Susan E

2016-03-01

Radiation-induced fibrosis is a common complication for patients following head and neck cancer treatment. This study presents a novel minimally invasive protocol for molecular study of fibrosis in the stromal tissues. Subjects with radiation-induced fibrosis in the head and neck who were at least 6 months post treatment received submental core needle biopsies, followed by molecular processing and quantification of gene expression for 14 select pro-inflammatory and pro-fibrotic genes. Control biopsies from the upper arm were obtained from the same subjects. Patients were followed up at 1 and 2 weeks to monitor for safety and adverse outcomes. Six subjects were enrolled and completed the study. No subjects experienced adverse outcomes or complication. An 18 gauge core biopsy needle with a 10 mm notch inserted for up to 60 seconds was needed. Subcutaneous tissue yielded 3 ng of RNA, amplified to 6 µg of cDNA, allowing for adequately sensitive quantitative polymerase chain reaction (qPCR) analysis of approximately 28 genes. This study demonstrates the safety and utility of a novel technique for the molecular study of fibrosis in head and neck cancer patients. Longitudinal studies of patients undergoing radiation therapy will allow for identification of molecular targets that contribute to the process of fibrosis in the head and neck. © The Author(s) 2015.

Initial Coupling of the RELAP-7 and PRONGHORN Applications

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

J. Ortensi; D. Andrs; A.A. Bingham

2012-10-01

Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations inmore » 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.« less

[Design and Analysis of CT High-speed Data Transmission Rotating Connector Ring System Retaining Ring].

PubMed

Pan, Li; Cao, Jujiang; Liu, Min; Fu, Weiwei

2017-11-30

High speed data transmission rotating connector system for signal high-speed transmission used in the fixed end and rotating end, it is one of the core component in the CT system. This paper involves structure design and analysis of the retaining ring in the CT high speed data transmission rotating connector system based on the principle of off-axis free space optical transmission. According to the problem of the actual engineering application of space limitations, optical fiber fixed and collimator installation location, we designed the structure of the retaining ring. Using the static analysis function of ANSYS Workbench, it verifies rationality and safety of the strength of retaining ring structure. And based on modal analysis function of ANSYS Workbench, it evaluates the effect of the retaining ring on the stability of the system date transmission, and provides theoretical basis for the feasibility of the structure in practical application.

Level 1 Tornado PRA for the High Flux Beam Reactor

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

Bozoki, G.E.; Conrad, C.S.

This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data,more » were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.« less

Self-criticism in therapist training: A grounded theory analysis.

PubMed

Kannan, Divya; Levitt, Heidi M

2017-03-01

The primary objective of this study is to engender an understanding of how therapists-in-training experience and cope with self-criticism in the context of their clinical training and therapy experiences. In this study, trainees were interviewed about their experience of self-criticism related to psychotherapy practice and these interviews were subjected to a grounded theory analysis generating a core self-critical process. The analysis highlighted the vulnerability of self-criticism in therapists' training experiences, especially when they related to balancing the "expert" role while maintaining authentic interactions with their clients. The results also described ways in which self-criticism is mitigated by a sense of interpersonal safety and the provision of clinical freedom and flexibility in therapists' training. The implications for future psychotherapy research and clinical training within clinical training environments are discussed.

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

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

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

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

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa

2002-07-01

A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled coremore » has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)« less

[Operating Room Nurses' Experiences of Securing for Patient Safety].

PubMed

Park, Kwang Ok; Kim, Jong Kyung; Kim, Myoung Sook

2015-10-01

This study was done to evaluate the experience of securing patient safety in hospital operating rooms. Experiential data were collected from 15 operating room nurses through in-depth interviews. The main question was "Could you describe your experience with patient safety in the operating room?". Qualitative data from the field and transcribed notes were analyzed using Strauss and Corbin's grounded theory methodology. The core category of experience with patient safety in the operating room was 'trying to maintain principles of patient safety during high-risk surgical procedures'. The participants used two interactional strategies: 'attempt continuous improvement', 'immersion in operation with sharing issues of patient safety'. The results indicate that the important factors for ensuring the safety of patients in the operating room are manpower, education, and a system for patient safety. Successful and safe surgery requires communication, teamwork and recognition of the importance of patient safety by the surgical team.

Evaluation of the Quality of Occupational Health and Safety Management Systems Based on Key Performance Indicators in Certified Organizations.

PubMed

Mohammadfam, Iraj; Kamalinia, Mojtaba; Momeni, Mansour; Golmohammadi, Rostam; Hamidi, Yadollah; Soltanian, Alireza

2017-06-01

Occupational Health and Safety Management Systems are becoming more widespread in organizations. Consequently, their effectiveness has become a core topic for researchers. This paper evaluates the performance of the Occupational Health and Safety Assessment Series 18001 specification in certified companies in Iran. The evaluation is based on a comparison of specific criteria and indictors related to occupational health and safety management practices in three certified and three noncertified companies. Findings indicate that the performance of certified companies with respect to occupational health and safety management practices is significantly better than that of noncertified companies. Occupational Health and Safety Assessment Series 18001-certified companies have a better level of occupational health and safety; this supports the argument that Occupational Health and Safety Management Systems play an important strategic role in health and safety in the workplace.

Software-Based Safety Systems in Space - Learning from other Domains

NASA Astrophysics Data System (ADS)

Klicker, M.; Putzer, H.

2012-01-01

Increasing complexity and new emerging capabilities for manned and unmanned missions have been the hallmark of the past decades of space exploration. One of the drivers in this process was the ever increasing use of software and software-intensive systems to implement system functions necessary to the capabilities needed. The course of technological evolution suggests that this development will continue well into the future with a number of challenges for the safety community some of which shall be discussed in this paper. The current state of the art reveals a number of problems with developing and assessing safety critical software which explains the reluctance of the space community to rely on software-based safety measures to mitigate hazards. Among others, usually lack of trustworthy evidence of software integrity in all foreseeable situations and the difficulties to integrate software in the traditional safety analysis framework are cited. Experience from other domains and recent developments in modern software development methodologies and verification techniques are analysed for the suitability for space systems and an avionics architectural framework (see STANAG 4626) for the implementation of safety critical software is proposed. This is shown to create among other features the possibility of numerous degradation modes enhancing overall system safety and interoperability of computerized space systems. It also potentially simplifies international cooperation on a technical level by introducing a higher degree of compatibility. As software safety cannot be tested or argued into a system in hindsight, the development process and especially the architecture chosen are essential to establish safety properties for the software used to implement safety functions. The core of the safety argument revolves around the separation of different functions and software modules from each other by minimal coupling of functions and credible separation mechanisms in the architecture combined with rigorous development methodologies for the software itself.

MAXimising Involvement in MUltiMorbidity (MAXIMUM) in primary care: protocol for an observation and interview study of patients, GPs and other care providers to identify ways of reducing patient safety failures

PubMed Central

Daker-White, Gavin; Hays, Rebecca; Esmail, Aneez; Minor, Brian; Barlow, Wendy; Brown, Benjamin; Blakeman, Thomas; Bower, Peter

2014-01-01

Introduction Increasing numbers of older people are living with multiple long-term health conditions but global healthcare systems and clinical guidelines have traditionally focused on the management of single conditions. Having two or more long-term conditions, or ‘multimorbidity’, is associated with a range of adverse consequences and poor outcomes and could put patients at increased risk of safety failures. Traditionally, most research into patient safety failures has explored hospital or inpatient settings. Much less is known about patient safety failures in primary care. Our core aims are to understand the mechanisms by which multimorbidity leads to safety failures, to explore the different ways in which patients and services respond (or fail to respond), and to identify opportunities for intervention. Methods and analysis We plan to undertake an applied ethnographic study of patients with multimorbidity. Patients’ interactions and environments, relevant to their healthcare, will be studied through observations, diary methods and semistructured interviews. A framework, based on previous studies, will be used to organise the collection and analysis of field notes, observations and other qualitative data. This framework includes the domains: access breakdowns, communication breakdowns, continuity of care errors, relationship breakdowns and technical errors. Ethics and dissemination Ethical approval was received from the National Health Service Research Ethics Committee for Wales. An individual case study approach is likely to be most fruitful for exploring the mechanisms by which multimorbidity leads to safety failures. A longitudinal and multiperspective approach will allow for the constant comparison of patient, carer and healthcare worker expectations and experiences related to the provision, integration and management of complex care. This data will be used to explore ways of engaging patients and carers more in their own care using shared decision-making, patient empowerment or other relevant models. PMID:25138807

The effectiveness of inking needle core prostate biopsies for preventing patient specimen identification errors: a technique to address Joint Commission patient safety goals in specialty laboratories.

PubMed

Raff, Lester J; Engel, George; Beck, Kenneth R; O'Brien, Andrea S; Bauer, Meagan E

2009-02-01

The elimination or reduction of medical errors has been a main focus of health care enterprises in the United States since the year 2000. Elimination of errors in patient and specimen identification is a key component of this focus and is the number one goal in the Joint Commission's 2008 National Patient Safety Goals Laboratory Services Program. To evaluate the effectiveness of using permanent inks to maintain specimen identity in sequentially submitted prostate needle biopsies. For a 12-month period, a grossing technician stained each prostate core with permanent ink developed for inking of pathology specimens. A different color was used for each patient, with all the prostate cores from all vials for a particular patient inked with the same color. Five colors were used sequentially: green, blue, yellow, orange, and black. The ink was diluted with distilled water to a consistency that allowed application of a thin, uniform coating of ink along the edges of the prostate core. The time required to ink patient specimens comprising different numbers of vials and prostate biopsies was timed. The number and type of inked specimen discrepancies were evaluated. The identified discrepancy rate for prostate biopsy patients was 0.13%. The discrepancy rate in terms of total number of prostate blocks was 0.014%. Diluted inks adhered to biopsy contours throughout tissue processing. The tissue showed no untoward reactions to the inks. Inking did not affect staining (histochemical or immunohistochemical) or pathologic evaluation. On average, inking prostate needle biopsies increases grossing time by 20%. Inking of all prostate core biopsies with colored inks, in sequential order, is an aid in maintaining specimen identity. It is a simple and effective method of addressing Joint Commission patient safety goals by maintaining specimen identity during processing of similar types of gross specimens. This technique may be applicable in other specialty laboratories and high-volume laboratories, where many similar tissue specimens are processed.

Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

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

Mark DeHart; William Skerjanc; Sean Morrell

2012-06-01

Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of themore » fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.« less

Impact of Micro-to Meso-scale Fractures on Sealing Behavior of Argillaceous Cap Rocks For CO 2 Sequestration

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

Evans, James

This multi-disciplinary project evaluated seal lithologies for the safety and security of long-term geosequestration of CO 2. We used integrated studies to provide qualitative risk for potential seal failure; we integrated data sets from outcrop, core, geochemical analysis, rock failure properties from mechanical testing, geophysical wireline log analysis, and geomechanical modeling to understand the effects of lithologic heterogeneity and changing mechanical properties have on the mechanical properties of the seal. The objectives of this study were to characterize cap rock seals using natural field analogs, available drillhole logging data and whole-rock core, geochemical and isotopic analyses. Rock deformation experiments weremore » carried out on collected samples to develop better models of risk estimation for potential cap rock seal failure. We also sampled variably faulted and fractured cap rocks to examine the impacts of mineralization and/or alteration on the mechanical properties. We compared CO 2 reacted systems to non-CO 2 reacted seal rock types to determine response of each to increased pore fluid pressures and potential for the creation of unintentional hydrofractures at depth.« less

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

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

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

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

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

Evaluating Reflectance Spectroscopy as a Method of Rapid Cryptotephra Identification using Component Analysis: Tephrochronology of the Lesser Antilles Arc

NASA Astrophysics Data System (ADS)

Fisher, E. A.

2015-12-01

The reactivation of Montserrat's South Soufrière-Soufrière Hills volcanic complex has impelled the creation of tephrochronologic records in the Lesser Antilles Arc in order to assess volcanic hazards to human safety. Developing an eruptive history of Montserrat by recording tephra layers preserved in marine sediment is hindered by the lack of a rapid, non-destructive method for detecting cryptotephra, tephra deposits invisible to the naked eye, in marine cores. Identifying cryptotephra is important because some cryptotephra layers represent primary tephra emplacement from small proximal eruptions, events that if excluded from a volcanic record could mischaracterize a volcano's eruptive frequency over time. VSWIR [0.4-2.5 μm] reflectance spectroscopy is a candidate for rapid, non-destructive cryptotephra detection in marine sediment cores because it can detect tephra in hemipelagic sediment using summary parameters sensitive to iron content and clay minerals (McCanta et al. 2014, AGU abstract OS53D-1086). Spectra from marine cores U1396C-1H-1A through U1396C-1H-5A, collected during International Ocean Discovery Program (IODP) mission 340, reveal 29 potential cryptotephra layers (McCanta et al. 2014, AGU abstract OS53D-1086). This study seeks to determine the effectiveness of reflectance spectroscopy at identifying cryptotephra by measuring the abundance of volcanic materials (i.e., glass shards/vesicular pumice and non-vesicular lava clasts) in these layers ( LeFriant et al. 2008; Cassidy et al. 2014). Component analysis was conducted on select core intervals with both cryptotephra-identifying peaks in reflectance parameters, and tephra-indicative peaks in core scanning XRF and magnetic susceptibility parameters (McCanta et al. 2014, AGU abstract OS53D-1086). Samples in this subset show abundances of non-vesicular lava and vesicular pumice clasts above expected background abundances, supporting the existence of cryptotephra at these locations (Fig. 1; LeFriant et al. 2008; Cassidy et al. 2014). This suggests that reflectance spectroscopy is an effective means of identifying cryptotephra in situ, and when employed in concert with other core scanning techniques could facilitate widespread rapid identification of cryptotephra in future tephrochronology studies.

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

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

IJ van Rooyen; SR Morrell; AE Wright

2014-10-01

Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizesmore » that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.« less

Direct Killing of Patients in Humanitarian Situations and Armed Conflicts: The Profession of Medicine Is Losing Its Meaning

PubMed Central

Asgary, Ramin

2015-01-01

During armed conflicts over the past several years, attacks on humanitarian workers and patients have increased, including the most recent overt killing of patients in their hospital beds in South Sudan and Central African Republic, and bombardments of hospitals in Iraq, Syria, and other countries. Direct attacks on patients inside hospitals, as well as social structural dynamics that undermine patient safety and security, are met with apparent indifference by international and medical communities. How can the medical profession remain silent and stand by while these factors render its core mission futile? In this article, I aim to shed light on this issue, and its implications for the future of the neutral and impartial provision of medical care; provide an analysis of underlying and contributing factors; discuss current international strategies; reflect on the responsibility of health providers; explore ways to strengthen our roles as physician advocates; and call for the medical profession to do more to protect medicine's core values. PMID:25646255

Population-level administration of AlcoholEdu for college: an ARIMA time-series analysis.

PubMed

Wyatt, Todd M; Dejong, William; Dixon, Elizabeth

2013-08-01

Autoregressive integrated moving averages (ARIMA) is a powerful analytic tool for conducting interrupted time-series analysis, yet it is rarely used in studies of public health campaigns or programs. This study demonstrated the use of ARIMA to assess AlcoholEdu for College, an online alcohol education course for first-year students, and other health and safety programs introduced at a moderate-size public university in the South. From 1992 to 2009, the university administered annual Core Alcohol and Drug Surveys to samples of undergraduates (Ns = 498 to 1032). AlcoholEdu and other health and safety programs that began during the study period were assessed through a series of quasi-experimental ARIMA analyses. Implementation of AlcoholEdu in 2004 was significantly associated with substantial decreases in alcohol consumption and alcohol- or drug-related negative consequences. These improvements were sustained over time as succeeding first-year classes took the course. Previous studies have shown that AlcoholEdu has an initial positive effect on students' alcohol use and associated negative consequences. This investigation suggests that these positive changes may be sustainable over time through yearly implementation of the course with first-year students. ARIMA time-series analysis holds great promise for investigating the effect of program and policy interventions to address alcohol- and drug-related problems on campus.

Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

NASA Astrophysics Data System (ADS)

Khatry, Jivan

Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

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

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

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

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

Validation of an Interdisciplinary Food Safety Curriculum Targeted at Middle School Students and Correlated to State Educational Standards

ERIC Educational Resources Information Center

Richards, Jennifer; Skolits, Gary; Burney, Janie; Pedigo, Ashley; Draughon, F. Ann

2008-01-01

Providing effective food safety education to young consumers is a national health priority to combat the nearly 76 million cases of foodborne illness in the United States annually. With the tremendous pressures on teachers for accountability in core subject areas, the focus of classrooms is on covering concepts that are tested on state performance…

Toward an African Maritime Economy: Empowering the African Union to Revolutionize the African Maritime Sector

DTIC Science & Technology

2011-01-01

maritime sector holistically, across its entire spectrum—improving safety and security, gover - nance, and industrial infrastructure and efficiency. There...includes the enabling elements of gover - nance, infrastructure, trade, safety, and security and plainly tells global partners where they can best...refugees, human rights, transparency, and accountability Infrastructure and Energy: transport (including maritime transport in its core function); tourism

Efficiency of static core turn-off in a system-on-a-chip with variation

DOEpatents

Cher, Chen-Yong; Coteus, Paul W; Gara, Alan; Kursun, Eren; Paulsen, David P; Schuelke, Brian A; Sheets, II, John E; Tian, Shurong

2013-10-29

A processor-implemented method for improving efficiency of a static core turn-off in a multi-core processor with variation, the method comprising: conducting via a simulation a turn-off analysis of the multi-core processor at the multi-core processor's design stage, wherein the turn-off analysis of the multi-core processor at the multi-core processor's design stage includes a first output corresponding to a first multi-core processor core to turn off; conducting a turn-off analysis of the multi-core processor at the multi-core processor's testing stage, wherein the turn-off analysis of the multi-core processor at the multi-core processor's testing stage includes a second output corresponding to a second multi-core processor core to turn off; comparing the first output and the second output to determine if the first output is referring to the same core to turn off as the second output; outputting a third output corresponding to the first multi-core processor core if the first output and the second output are both referring to the same core to turn off.

Building the Core Architecture of a Multiagent System Product Line: With an example from a future NASA Mission

NASA Technical Reports Server (NTRS)

Pena, Joaquin; Hinchey, Michael G.; Ruiz-Cortes, Antonio

2006-01-01

The field of Software Product Lines (SPL) emphasizes building a core architecture for a family of software products from which concrete products can be derived rapidly. This helps to reduce time-to-market, costs, etc., and can result in improved software quality and safety. Current AOSE methodologies are concerned with developing a single Multiagent System. We propose an initial approach to developing the core architecture of a Multiagent Systems Product Line (MAS-PL), exemplifying our approach with reference to a concept NASA mission based on multiagent technology.

Employer Health and Productivity Roadmap™ strategy.

PubMed

Parkinson, Michael D

2013-12-01

The National Institute for Occupational Safety and Health Total Worker Health™ Program defines essential elements of an integrated health protection and health promotion model to improve the health, safety, and performance of employers and employees. The lack of a clear strategy to address the core drivers of poor health, excessive medical costs, and lost productivity has deterred a comprehensive, integrated, and proactive approach to meet these challenges. The Employer Health and Productivity Roadmap™, comprising six interrelated and integrated core elements, creates a framework of shared accountability for both employers and their health and productivity partners to implement and monitor actionable measures that improve health, maximize productivity, and reduce excessive costs. The strategy is most effective when linked to a financially incentivized health management program or consumer-directed health plan insurance benefit design.

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

[Legal development of consumer protection from the Federal Office of Consumer Protection and Food Safety standpoint].

PubMed

Püster, M

2010-06-01

Ten years after publication of the White Paper on Food Safety, health consumer protection has made significant progress and, today, is a key field in politics at both the European and German levels. In addition to the protection of health and security of consumers, consumer information has become a core element of consumer protection for the Federal Office of Consumer Protection and Food Safety (Bundesamt für Verbraucherschutz and Lebensmittelsicherheit, BVL). State authorities are provided with new means of communication and interaction with consumers.

Commercial grade item (CGI) dedication of MDR relays for nuclear safety related applications

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

Das, R.K.; Julka, A.; Modi, G.

1994-08-01

MDR relays manufactured by Potter and Brumfield (P and B) have been used in various safety related applications in commercial nuclear power plants. These include emergency safety features (ESF) actuation systems, emergency core cooling systems (ECCS) actuation, and reactor protection systems. The MDR relays manufactured prior to May 1990 showed signs of generic failure due to corrosion and outgassing of coil varnish. P and B has made design changes to correct these problems in relays manufactured after May 1990. However, P and B does not manufacture the relays under any 10CFR50 Appendix B quality assurance (QA) program. They manufacture themore » relays under their commercial QA program and supply these as commercial grade items. This necessitates CGI Dedication of these relays for use in nuclear-safety-related applications. This paper presents a CGI dedication program that has been used to dedicate the MDR relays manufactured after May 1990. The program is in compliance with current Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) guidelines and applicable industry standards; it specifies the critical characteristics of the relays, provides the tests and analysis required to verify the critical characteristics, the acceptance criteria for the test results, performs source verification to qualify P and B for its control of the critical characteristics, and provides documentation. The program provides reasonable assurance that the new MDR relays will perform their intended safety functions.« less

Patient safety ward round checklist via an electronic app: implications for harm prevention.

PubMed

Keller, C; Arsenault, S; Lamothe, M; Bostan, S R; O'Donnell, R; Harbison, J; Doherty, C P

2017-11-06

Patient safety is a value at the core of modern healthcare. Though awareness in the medical community is growing, implementing systematic approaches similar to those used in other high reliability industries is proving difficult. The aim of this research was twofold, to establish a baseline for patient safety practices on routine ward rounds and to test the feasibility of implementing an electronic patient safety checklist application. Two research teams were formed; one auditing a medical team to establish a procedural baseline of "usual care" practice and an intervention team concurrently was enforcing the implementation of the checklist. The checklist was comprised of eight standard clinical practice items. The program was conducted over a 2-week period and 1 month later, a retrospective analysis of patient charts was conducted using a global trigger tool to determine variance between the experimental groups. Finally, feedback from the physician participants was considered. The results demonstrated a statistically significant difference on five variables of a total of 16. The auditing team observed low adherence to patient identification (0.0%), hand decontamination (5.5%), and presence of nurse on ward rounds (6.8%). Physician feedback was generally positive. The baseline audit demonstrated significant practice bias on daily ward rounds which tended to omit several key-proven patient safety practices such as prompting hand decontamination and obtaining up to date reports from nursing staff. Results of the intervention arm demonstrate the feasibility of using the Checklist App on daily ward rounds.

Patient Safety Leadership WalkRounds.

PubMed

Frankel, Allan; Graydon-Baker, Erin; Neppl, Camilla; Simmonds, Terri; Gustafson, Michael; Gandhi, Tejal K

2003-01-01

In the WalkRounds concept, a core group, which includes the senior executives and/or vice presidents, conducts weekly visits to different areas of the hospital. The group, joined by one or two nurses in the area and other available staff, asks specific questions about adverse events or near misses and about the factors or systems issues that led to these events. ANALYSIS OF EVENTS: Events in the Walkrounds are entered into a database and classified according to the contributing factors. The data are aggregated by contributing factors and priority scores to highlight the root issues. The priority scores are used to determine QI pilots and make best use of limited resources. Executives are surveyed quarterly about actions they have taken as a direct result of WalkRounds and are asked what they have learned from the rounds. As of September 2002, 47 Patient Safety Leadership WalkRounds visited a total of 48 different areas of the hospital, with 432 individual comments. The WalkRounds require not only knowledgeable and invested senior leadership but also a well-organized support structure. Quality and safety personnel are needed to collect data and maintain a database of confidential information, evaluate the data from a systems approach, and delineate systems-based actions to improve care delivery. Comments of frontline clinicians and executives suggested that WalkRounds helps educate leadership and frontline staff in patient safety concepts and will lead to cultural changes, as manifested in more open discussion of adverse events and an improved rate of safety-based changes.

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.

Upgrade of Irradiation Test Capability of the Experimental Fast Reactor Joyo

NASA Astrophysics Data System (ADS)

Sekine, Takashi; Aoyama, Takafumi; Suzuki, Soju; Yamashita, Yoshioki

2003-06-01

The JOYO MK-II core was operated from 1983 to 2000 as fast neutron irradiation bed. In order to meet various requirements for irradiation tests for development of FBRs, the JOYO upgrading project named MK-III program was initiated. The irradiation capability in the MK-III core will be about four times larger than that of the MK-II core. Advanced irradiation test subassemblies such as capsule type subassembly and on-line instrumentation rig are planned. As an innovative reactor safety system, the irradiation test of Self-Actuated Shutdown System (SASS) will be conducted. In order to improve the accuracy of neutron fluence, the core management code system was upgraded, and the Monte Carlo code and Helium Accumulation Fluence Monitor (HAFM) were applied. The MK-III core is planned to achieve initial criticality in July 2003.

76 FR 5651 - Practice and Procedure; Amendment of CORES Registration System

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-01

..., including the Antenna Structure Registration System (``ASR'') (managed by the Commission's Wireless... Wireless Telecommunications Bureau and the Public Safety and Homeland Security Bureau). Among other things...

Inspection of the Math Model Tools for On-Orbit Assessment of Impact Damage Report

NASA Technical Reports Server (NTRS)

Harris, Charles E.; Raju, Ivatury S.; Piascik, Robert S> ; KramerWhite, Julie A.; KramerWhite, Julie A.; Labbe, Steve G.; Rotter, Hank A.

2007-01-01

In Spring of 2005, the NASA Engineering Safety Center (NESC) was engaged by the Space Shuttle Program (SSP) to peer review the suite of analytical tools being developed to support the determination of impact and damage tolerance of the Orbiter Thermal Protection Systems (TPS). The NESC formed an independent review team with the core disciplines of materials, flight sciences, structures, mechanical analysis and thermal analysis. The Math Model Tools reviewed included damage prediction and stress analysis, aeroheating analysis, and thermal analysis tools. Some tools are physics-based and other tools are empirically-derived. Each tool was created for a specific use and timeframe, including certification, real-time pre-launch assessments. In addition, the tools are used together in an integrated strategy for assessing the ramifications of impact damage to tile and RCC. The NESC teams conducted a peer review of the engineering data package for each Math Model Tool. This report contains the summary of the team observations and recommendations from these reviews.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Hazard analysis and critical control point systems in the United States Department of Agriculture regulatory policy.

PubMed

Billy, T J; Wachsmuth, I K

1997-08-01

Recent outbreaks of foodborne illness and studies by expert groups have established the need for fundamental change in the United States meat and poultry inspection programme to reduce the risk of foodborne illness. The Food Safety and Inspection Service (FSIS) of the United States Department of Agriculture (USDA) has embarked on a broad effort to bring about such change, with particular emphasis on the reduction of pathogenic micro-organisms in raw meat and poultry products. The publication on 25 July 1996 of the Final Rule on pathogen reduction and hazard analysis and critical control point (HACCP) systems was a major milestone in the FSIS strategy for change. The Final Rule provides a framework for change and clarifies the respective roles of industry and government in ensuring the safety of meat and poultry products. With the implementation of this Final Rule underway, the FSIS has been exploring ways in which slaughter inspection carried out under an HACCP-based system can be changed so that food safety risks are addressed more adequately and the allocation of inspection resources is improved further. In addition, the FSIS is broadening the focus of food safety activities to extend beyond slaughter and processing plants by working with industry, academia and other government agencies. Such co-operation should lead to the development of measures to improve food safety before animals reach the slaughter plant and after products leave the inspected establishment for distribution to the retail level. For the future, the FSIS believes that quantitative risk assessments will be at the core of food safety activities. Risk assessments provide the most effective means of identifying how specific pathogens and other hazards may be encountered throughout the farm-to-table chain and of measuring the potential impact of various interventions. In addition, these assessments will be used in the development and evaluation of HACCP systems. The FSIS is currently conducting a quantitative risk assessment for eggs, and several surveys and studies are being performed to supply data needed to conduct other risk assessments. The FSIS has established a food safety research agenda which will fill data gaps.

[Patient safety in education and training of healthcare professionals in Germany].

PubMed

Hoffmann, Barbara; Siebert, H; Euteneier, A

2015-01-01

In order to improve patient safety, healthcare professionals who care for patients directly or indirectly are required to possess specific knowledge and skills. Patient safety education is not or only poorly represented in education and examination regulations of healthcare professionals in Germany; therefore, it is only practiced rarely and on a voluntary basis. Meanwhile, several training curricula and concepts have been developed in the past 10 years internationally and recently in Germany, too. Based on these concepts the German Coalition for Patient Safety developed a catalogue of core competencies required for safety in patient care. This catalogue will serve as an important orientation when patient safety is to be implemented as a subject of professional education in Germany in the future. Moreover, teaching staff has to be trained and educational and training activities have to be evaluated. Patient safety education and training for (undergraduate) healthcare professional will require capital investment.

How pleasant sounds promote and annoying sounds impede health: a cognitive approach.

PubMed

Andringa, Tjeerd C; Lanser, J Jolie L

2013-04-08

This theoretical paper addresses the cognitive functions via which quiet and in general pleasurable sounds promote and annoying sounds impede health. The article comprises a literature analysis and an interpretation of how the bidirectional influence of appraising the environment and the feelings of the perceiver can be understood in terms of core affect and motivation. This conceptual basis allows the formulation of a detailed cognitive model describing how sonic content, related to indicators of safety and danger, either allows full freedom over mind-states or forces the activation of a vigilance function with associated arousal. The model leads to a number of detailed predictions that can be used to provide existing soundscape approaches with a solid cognitive science foundation that may lead to novel approaches to soundscape design. These will take into account that louder sounds typically contribute to distal situational awareness while subtle environmental sounds provide proximal situational awareness. The role of safety indicators, mediated by proximal situational awareness and subtle sounds, should become more important in future soundscape research.

Comprehensive in vitro Proarrhythmia Assay (CiPA): Pending issues for successful validation and implementation.

PubMed

Cavero, Icilio; Guillon, Jean-Michel; Ballet, Veronique; Clements, Mike; Gerbeau, Jean-Frédéric; Holzgrefe, Henry

2016-01-01

The Comprehensive in vitro Proarrhythmia Assay (CiPA) is a nonclinical Safety Pharmacology paradigm for discovering electrophysiological mechanisms that are likely to confer proarrhythmic liability to drug candidates intended for human use. Key talks delivered at the 'CiPA on my mind' session, held during the 2015 Annual Meeting of the Safety Pharmacology Society (SPS), are summarized. Issues and potential solutions relating to crucial constituents [e.g., biological materials (ion channels and pluripotent stem cell-derived cardiomyocytes), study platforms, drug solutions, and data analysis] of CiPA core assays are critically examined. In order to advance the CiPA paradigm from the current testing and validation stages to a research and regulatory drug development strategy, systematic guidance by CiPA stakeholders is necessary to expedite solutions to pending and newly arising issues. Once a study protocol is proved to yield robust and reproducible results within and across laboratories, it can be implemented as qualified regulatory procedure. Copyright © 2016 Elsevier Inc. All rights reserved.

Type C investigation of electrical fabrication projects in ICF Kaiser shops

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

Huckfeldt, R.A.

1995-06-01

A Type C Investigation Board was convened to investigate an electrical miswiring problem found during the operation of the electrical distribution trailer for the TWRS Rotary Mode Core Sampling Truck {number_sign}2. The trailer was designed by WHC and fabricated ICF KH on site for use in the Characterization Program. This problem resulted in a serious safety hazard since the support truck frame/chassis became electrically energized. This final report provides results of the ``Type C Investigation, Electrical Fabrication Projects in ICF KH Shops, June, 1995.`` It contains the investigation scope, executive summary, relevant facts, analysis, conclusions and corrective actions. DOE Ordermore » 5484.1, ``Environmental Protection, Safety and Health Protection Information Reporting Requirements,`` was followed in preparation of this report. Because the incident was electrical in nature and involved both Westinghouse Hanford Company and ICF Kaiser Hanford organizations, the board included members from both contractors and members with considerable electrical expertise.« 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

AACE: an innovative partnership to enhance aircraft safety

NASA Astrophysics Data System (ADS)

Shurtleff, William W.

1999-01-01

The Federal Aviation Administration established the Airworthiness Assurance Center of Excellence (AACE) in September 1997, through a cooperative agreement grant with Iowa State University (ISU) and The Ohio State University (OSU). A technical support contract with the Center is now in place as well. Initially the Center has five areas of concentration supporting advances in airworthiness assurance. These are 1. Maintenance, inspection, and repair, 2. Propulsion and fuel systems safety, 3. Crashworthiness, 4. Advanced materials, and 5. Landing gear systems performance and safety. AACE has nine core members who provide guidance to the Program Management Office at ISU/OSU through a Board of Directors. The core members are: Arizona State University, Iowa State University, Northwestern University, The Ohio State University, University of Dayton, University of Maryland, University of California - Los Angeles, Wichita State University, and Sandia National Laboratories. The organization also includes numerous academic affiliates, industry partners, government laboratories and other organizations. The Center now has over thirty technical projects supporting technical advances in airworthiness assurance. All these projects have industry guidance and support. This paper discusses the current technical program of the center and the highlights of the five-year plan for technical work. Also included is a description of the factors that make the Center an innovative partnership to promote aircraft safety.

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

Initiating Event Analysis of a Lithium Fluoride Thorium Reactor

NASA Astrophysics Data System (ADS)

Geraci, Nicholas Charles

The primary purpose of this study is to perform an Initiating Event Analysis for a Lithium Fluoride Thorium Reactor (LFTR) as the first step of a Probabilistic Safety Assessment (PSA). The major objective of the research is to compile a list of key initiating events capable of resulting in failure of safety systems and release of radioactive material from the LFTR. Due to the complex interactions between engineering design, component reliability and human reliability, probabilistic safety assessments are most useful when the scope is limited to a single reactor plant. Thus, this thesis will study the LFTR design proposed by Flibe Energy. An October 2015 Electric Power Research Institute report on the Flibe Energy LFTR asked "what-if?" questions of subject matter experts and compiled a list of key hazards with the most significant consequences to the safety or integrity of the LFTR. The potential exists for unforeseen hazards to pose additional risk for the LFTR, but the scope of this thesis is limited to evaluation of those key hazards already identified by Flibe Energy. These key hazards are the starting point for the Initiating Event Analysis performed in this thesis. Engineering evaluation and technical study of the plant using a literature review and comparison to reference technology revealed four hazards with high potential to cause reactor core damage. To determine the initiating events resulting in realization of these four hazards, reference was made to previous PSAs and existing NRC and EPRI initiating event lists. Finally, fault tree and event tree analyses were conducted, completing the logical classification of initiating events. Results are qualitative as opposed to quantitative due to the early stages of system design descriptions and lack of operating experience or data for the LFTR. In summary, this thesis analyzes initiating events using previous research and inductive and deductive reasoning through traditional risk management techniques to arrive at a list of key initiating events that can be used to address vulnerabilities during the design phases of LFTR development.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

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

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

Ultrasonic approach to the synthesis of HMX@TATB core-shell microparticles with improved mechanical sensitivity.

PubMed

Huang, Bing; Hao, Xiaofei; Zhang, Haobin; Yang, Zhijian; Ma, Zhigang; Li, Hongzhen; Nie, Fude; Huang, Hui

2014-07-01

To improve the safety of sensitive explosive HMX while maintaining explosion performance, a moderately powerful but insensitive explosive TATB was used to coat HMX microparticles via a facile ultrasonic method. By using Estane as surface modifier and nano-sized TATB as the shell layer, the HMX@TATB core-shell microparticles with a monodisperse size and compact shell structure were successfully constructed. Both scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS) results confirmed the formation of perfect core-shell structured composites. Based on a systematic and comparative study of the effect of experimental conditions, a possible formation mechanism of core-shell structure was proposed in detail. Moreover, the perfect core-shell HMX@TATB microparticles exhibited a unique thermal behavior and significantly improved mechanical sensitivity compared with that of the physical mixture. Copyright © 2014 Elsevier B.V. All rights reserved.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

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

Ilas, Germina; Chandler, David; Ade, Brian J

2015-03-01

A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the designmore » of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.« less

Switching patients with acromegaly from octreotide to pasireotide improves biochemical control: crossover extension to a randomized, double-blind, Phase III study.

PubMed

Bronstein, Marcello D; Fleseriu, Maria; Neggers, Sebastian; Colao, Annamaria; Sheppard, Michael; Gu, Feng; Shen, Chiung-Chyi; Gadelha, Mônica; Farrall, Andrew J; Hermosillo Reséndiz, Karina; Ruffin, Matthieu; Chen, YinMiao; Freda, Pamela

2016-04-02

Many patients with acromegaly do not achieve biochemical control with first-generation somatostatin analogues. A large, multicenter, randomized, Phase III core study demonstrated that pasireotide LAR had significantly superior efficacy over octreotide LAR. This analysis explores the efficacy and safety of switching therapeutic arms in inadequately controlled patients during a 12-month crossover extension. Patients with inadequate biochemical control (GH ≥2.5 μg/L and/or IGF-1 > ULN) at end of core study (month 12) were eligible to switch to pasireotide LAR 40 mg/28 days (n = 81) or octreotide LAR 20 mg/28 days (n = 38). One dose escalation to pasireotide LAR 60 mg/28 days or octreotide LAR 30 mg/28 days was permitted, but not mandatory, at month 17 or 20. Twelve months after crossover, 17.3 % of pasireotide LAR and 0 % of octreotide LAR patients achieved GH <2.5 μg/L and normal IGF-1 (main outcome measure); 27.2 and 5.3 % of pasireotide LAR and octreotide LAR patients achieved normal IGF-1, respectively; 44.4 and 23.7 % of pasireotide LAR and octreotide LAR patients achieved GH <2.5 μg/L, respectively. Mean (±SD) tumor volume further decreased from the end of the core study by 25 % (±25) and 18 % (±28); 54.3 % of pasireotide LAR and 42.3 % of octreotide LAR patients achieved significant (≥20 %) tumor volume reduction during the extension. The safety profile of pasireotide LAR was similar to that of octreotide LAR, with the exception of the frequency and degree of hyperglycemia-related adverse events. Pasireotide LAR is a promising treatment option for patients with acromegaly inadequately controlled with the first-generation somatostatin analogue octreotide LAR. clinicaltrials.gov, NCT00600886 . Registered 14 January 2008.

The Role of Affect in Cross-Cultural Competence

DTIC Science & Technology

2012-04-26

4.56 2.58 .70 .89 Food Affect (pre) 1.00 5.00 3.10 .70 .91 Food Affect (post) 1.00 5.00 3.22 1.00 .95 Food Safety (pre) 1.57 5.00 3.35 .82 Food ...controlled for  Looked at affect in terms of food affect and food safety Hypotheses and Results (cont.) • Hypothesis 2a*: Disgust sensitivity will be...05 • Food Safety , r(96)=-.13, n.s. • Disgust Sensitivity, r(96)=-.16, n.s. • Contamination, r(96)=-.24, p<.05 • Core Disgust, r(96)=-.20, p=.05

[A study on the relationship between point mutation in pre-core region G1896A of hepatitis B virus and safety of breast feeding].

PubMed

Lu, Yin-ping; Cao, Wei; Hong, Mei; Zhu, Jian-fang; Liu, Zhao; Yang, Dong-liang

2008-10-01

To investigate the relationship between pre-core G1896A point mutation of hepatitis B virus (HBV) and safety of breast feeding. Serum and breast milk samples were collected from 62 pregnant women of HBV DNA positive/HBeAg negative. PCR-solid phase hybridization was used to detect the point mutation in pre-core region G1896A of HBV from pregnant women, and HBV DNA loads in sera and breast milk were determined by fluorescence quantitative PCR (FQ-PCR). The prevalence of point mutation was 61.3% (38/62) in 62 pregnant women with HBsAg positive/HBeAg negative. The positive rate of HBV DNA in breast milk of group with point mutation (28.9%) was similar to that of group without mutation (29.2%, chi2=0.0003, P>0.05). However, The positive rate of HBV DNA in breast milk of group with high HBV loads (56.0%) was significantly higher than that of group with low HBV loads (10.8%, chi2=14.79, P<0.01). The point mutation in pre-core region G1896A of HBV dose not affect the positive rate of HBV DNA in breast milk and higher HBV DNA loads in serum of pregnant women might increase the risk of mother-infant transmission.

Preliminary Structural Sizing and Alternative Material Trade Study of CEV Crew Module

NASA Technical Reports Server (NTRS)

Bednarcyk, Brett A.; Arnold, Steve M.; Collier, Craig S.; Yarrington, Phillip W.

2007-01-01

This paper presents the results of a preliminary structural sizing and alternate material trade study for NASA s Crew Exploration Vehicle (CEV) Crew Module (CM). This critical CEV component will house the astronauts during ascent, docking with the International Space Station, reentry, and landing. The alternate material design study considers three materials beyond the standard metallic (aluminum alloy) design that resulted from an earlier NASA Smart Buyer Team analysis. These materials are graphite/epoxy composite laminates, discontinuously reinforced SiC/Al (DRA) composites, and a novel integrated panel material/concept known as WebCore. Using the HyperSizer (Collier Research and Development Corporation) structural sizing software and NASTRAN finite element analysis code, a comparison is made among these materials for the three composite CM concepts considered by the 2006 NASA Engineering and Safety Center Composite Crew Module project.

Core continuing professional development (CPD) topics for the European dentist.

PubMed

Bailey, S; Bullock, A; Cowpe, J; Barnes, E; Thomas, H; Thomas, R; Kavadella, A; Kossioni, A; Karaharju-Suvanto, T; Suomalainen, K; Kersten, H; Povel, E; Giles, M; Walmsley, A D; Soboleva, U; Liepa, A; Akota, I

2013-05-01

In the context of free movement, EU-citizens need assurance that dental practitioners providing their care have a degree/license to practice that meets EU-standards and that they maintain their knowledge and skills through ongoing education. One aim of the 'DentCPD' project (HYPERLINK 'http://www.dentcpd.org' www.dentcpd.org) was to identify and agree essential CPD requirements for EU dentists. This paper reports the consensus process and outcomes. Agreement on core components of CPD was achieved through a three stage process: an online survey of dental educators' (n = 143) views on compulsory topics; a paper-based questionnaire to practitioners (n = 411); leading to a proposal discussed at the Association for Dental Education (ADEE) 2011 Lifelong Learning special interest group (SIG). From the online survey and practitioner questionnaire, high levels of agreement were achieved for medical emergencies (89%), infection control (79%) and the medically compromised patient (71%). The SIG (34 attendees from 16 countries) concluded that these three CPD topics plus radiation protection should be core-compulsory and three CPD topics should be core-recommended (health and safety, pain management, and safeguarding children & vulnerable adults). They also agreed that the teaching of all topics should be underpinned by evidence-based dentistry. Building four core topics into CPD requirements and making quality-approved education and training available will ensure that all dentists have up-to-date knowledge and skills in topic areas of direct relevance to patient safety. In turn, this will contribute to patients having access to comparably high standards of oral health care across Europe. © 2013 John Wiley & Sons A/S.

Two sides of the safety coin?: How patient engagement and safety climate jointly affect error occurrence in hospital units.

PubMed

Schiffinger, Michael; Latzke, Markus; Steyrer, Johannes

2016-01-01

Safety climate (SC) and more recently patient engagement (PE) have been identified as potential determinants of patient safety, but conceptual and empirical studies combining both are lacking. On the basis of extant theories and concepts in safety research, this study investigates the effect of PE in conjunction with SC on perceived error occurrence (pEO) in hospitals, controlling for various staff-, patient-, and hospital-related variables as well as the amount of stress and (lack of) organizational support experienced by staff. Besides the main effects of PE and SC on error occurrence, their interaction is examined, too. In 66 hospital units, 4,345 patients assessed the degree of PE, and 811 staff assessed SC and pEO. PE was measured with a new instrument, capturing its core elements according to a recent literature review: Information Provision (both active and passive) and Activation and Collaboration. SC and pEO were measured with validated German-language questionnaires. Besides standard regression and correlational analyses, partial least squares analysis was employed to model the main and interaction effects of PE and SC on pEO, also controlling for stress and (lack of) support perceived by staff, various staff and patient attributes, and potential single-source bias. Both PE and SC are associated with lower pEO, to a similar extent. The joint effect of these predictors suggests a substitution rather than mutually reinforcing interaction. Accounting for control variables and/or potential single-source bias slightly attenuates some effects without altering the results. Ignoring PE potentially amounts to forgoing a potential source of additional safety. On the other hand, despite the abovementioned substitution effect and conjectures of SC being inert, PE should not be considered as a replacement for SC.

Real-world daptomycin use across wide geographical regions: results from a pooled analysis of CORE and EU-CORE.

PubMed

Seaton, R Andrew; Gonzalez-Ruiz, Armando; Cleveland, Kerry O; Couch, Kimberly A; Pathan, Rashidkhan; Hamed, Kamal

2016-03-15

Pooled data from two large registries, Cubicin(®) Outcomes Registry and Experience (CORE; USA) and European Cubicin(®) Outcomes Registry and Experience (EU-CORE; Europe, Latin America, and Asia), were analyzed to determine the characteristics and clinical outcomes of daptomycin therapy in patients with Gram-positive infections across wide geographical regions. Patients receiving at least one dose of daptomycin between 2004 and 2012 for the treatment of Gram-positive infections were included. Clinical success was defined as an outcome of 'cured' or 'improved'. Post-treatment follow-up data were collected for a subset of patients (CORE: osteomyelitis and orthopedic foreign body device infection; EU-CORE: endocarditis, intracardiac/intravascular device infection, osteomyelitis, and orthopedic device infection). Safety was assessed for up to 30 days after daptomycin treatment. In 11,557 patients (CORE, 5482; EU-CORE, 6075) treated with daptomycin (median age, 62 [range, 1-103] years), the most frequent underlying conditions were cardiovascular disease (54.7 %) and diabetes mellitus (28.0 %). The most commonly treated primary infections were complicated skin and soft tissue infection (cSSTI; 31.2 %) and bacteremia (21.8 %). The overall clinical success rate was 77.2 % (uncomplicated SSTI, 88.3 %; cSSTI, 81.0 %; osteomyelitis, 77.7 %; foreign body/prosthetic infection (FBPI), 75.9 %; endocarditis, 75.4 %; and bacteremia, 69.5 %). The clinical success rate was 79.1 % in patients with Staphylococcus aureus infections (MRSA, 78.1 %). An increasing trend of high-dose daptomycin (>6 mg/kg/day) prescribing pattern was observed over time. Clinical success rates were higher with high-dose daptomycin treatment for endocarditis and FBPI. Adverse events (AEs) and serious AEs possibly related to daptomycin therapy were reported in 628 (5.4 %) and 133 (1.2 %) patients, respectively. The real-world data showed that daptomycin was effective and safe in the treatment of various Gram-positive infections, including those caused by resistant pathogens, across wide geographical regions.

Dual-core optical fiber based strain sensor for remote sensing in hard-to-reach areas

NASA Astrophysics Data System (ADS)

MÄ kowska, Anna; Szostkiewicz, Łukasz; Kołakowska, Agnieszka; Budnicki, Dawid; Bieńkowska, Beata; Ostrowski, Łukasz; Murawski, Michał; Napierała, Marek; Mergo, Paweł; Nasiłowski, Tomasz

2017-10-01

We present research on optical fiber sensors based on microstructured multi-core fiber. Elaborated sensor can be advantageously used in hard-to-reach areas by taking advantage of the fact, that optical fibers can play both the role of sensing elements and they can realize signal delivery. By using the sensor, it is possible to increase the level of the safety in the explosive endangered areas, e.g. in mine-like objects. As a base for the strain remote sensor we use dual-core fibers. The multi-core fibers possess a characteristic parameter called crosstalk, which is a measure of the amount of signal which can pass to the adjacent core. The strain-sensitive area is made by creating the tapered section, in which the level of crosstalk is changed. Due to this fact, we present broadened conception of fiber optic sensor designing. Strain measurement is realized thanks to the fact, that depending on the strain applied, the power distribution between the cores of dual-core fibers changes. Principle of operation allows realization of measurements both in wavelength and power domain.

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Study Progress of Physiological Responses in High Temperature Environment

NASA Astrophysics Data System (ADS)

Li, K.; Zheng, G. Z.; Bu, W. T.; Wang, Y. J.; Lu, Y. Z.

2017-10-01

Certain workers are exposed to high temperatures for a long time. Heat stress will result in a series of physiological responses, and cause adverse effects on the health and safety of workers. This paper summarizes the physiological changes of cardiovascular system, core temperature, skin temperature, water-electrolyte metabolism, alimentary system, neuroendocrine system, reaction time and thermal fatigue in high temperature environments. It can provide a theoretical guidance for labor safety in high temperature environment.

Identifying core herbal treatments for children with asthma: implication from a chinese herbal medicine database in taiwan.

PubMed

Chen, Hsing-Yu; Lin, Yi-Hsuan; Thien, Peck-Foong; Chang, Shih-Chieh; Chen, Yu-Chun; Lo, Su-Shun; Yang, Sien-Hung; Chen, Jiun-Liang

2013-01-01

Asthma is one of the most common allergic respiratory diseases around the world and places great burden on medical payment. Chinese herbal medicine (CHM) is commonly used for Taiwanese children to control diseases. The aim of this study is to analyze the CHM prescriptions for asthmatic children by using a nationwide clinical database. The National Health Insurance Research Database (NHIRD) was used to perform this study. Medical records from 1997 to 2009 with diagnosis with asthma made for children aged 6 to 18 were included into the analysis. Association rule mining and social network analysis were used to analyze the prevalence of single CHM and its combinations. Ma-Xing-Gan-Shi-Tang (MXGST) was the most commonly used herbal formula (HF) (20.2% of all prescriptions), followed by Xiao-Qing-Long-Tang (13.1%) and Xing-Su-San (12.8%). Zhe Bei Mu is the most frequently used single herb (SH) (14.6%), followed by Xing Ren (10.7%). MXGST was commonly used with Zhe Bei Mu (3.5%) and other single herbs capable of dispelling phlegm. Besides, MXGST was the core formula to relieve asthma. Further studies about efficacy and drug safety are needed for the CHM commonly used for asthma based on the result of this study.

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

Rouxelin, Pascal Nicolas; Strydom, Gerhard

Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented bymore » the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise II 1a. The steady state core calculations were simulated with the INL coupled-code system known as the Parallel and Highly Innovative Simulation for INL Code System (PHISICS) and the system thermal-hydraulics code known as the Reactor Excursion and Leak Analysis Program (RELAP) 5 3D using the nuclear data libraries previously generated with NEWT. It was observed that significant differences in terms of multiplication factor and neutron flux exist between the various permutations of the Phase I super-cell lattice calculations. The use of these cross section libraries only leads to minor changes in the Phase II core simulation results for fresh fuel but shows significantly larger discrepancies for spent fuel cores. Furthermore, large incongruities were found between the SCALE NEWT and KENO VI results for the super cells, and while some trends could be identified, a final conclusion on this issue could not yet be reached. This report will be revised in mid 2016 with more detailed analyses of the super-cell problems and their effects on the core models, using the latest version of SCALE (6.2). The super-cell models seem to show substantial improvements in terms of neutron flux as compared to single-block models, particularly at thermal energies.« less

46 CFR 153.214 - Personnel emergency and safety equipment.

Code of Federal Regulations, 2010 CFR

2010-10-01

...-propelled ship must have the following: (a) Two stretchers or wire baskets complete with equipment for... Inspection or Certificate of Compliance. (3) A steel-cored lifeline with harness. (4) An explosion-proof lamp...

46 CFR 153.214 - Personnel emergency and safety equipment.

Code of Federal Regulations, 2011 CFR

2011-10-01

...-propelled ship must have the following: (a) Two stretchers or wire baskets complete with equipment for... Inspection or Certificate of Compliance. (3) A steel-cored lifeline with harness. (4) An explosion-proof lamp...

78 FR 47010 - Proposed Safety Evaluation for Plant-Specific

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-02

... to the existing SR on the reactor core isolation cooling system to maintain consistency within the... TS Bases are revised to reflect the change to the SRs. The proposed change captures the on-going...

46 CFR 153.214 - Personnel emergency and safety equipment.

Code of Federal Regulations, 2012 CFR

2012-10-01

...-propelled ship must have the following: (a) Two stretchers or wire baskets complete with equipment for... Inspection or Certificate of Compliance. (3) A steel-cored lifeline with harness. (4) An explosion-proof lamp...

46 CFR 153.214 - Personnel emergency and safety equipment.

Code of Federal Regulations, 2013 CFR

2013-10-01

...-propelled ship must have the following: (a) Two stretchers or wire baskets complete with equipment for... Inspection or Certificate of Compliance. (3) A steel-cored lifeline with harness. (4) An explosion-proof lamp...

46 CFR 153.214 - Personnel emergency and safety equipment.

Code of Federal Regulations, 2014 CFR

2014-10-01

...-propelled ship must have the following: (a) Two stretchers or wire baskets complete with equipment for... Inspection or Certificate of Compliance. (3) A steel-cored lifeline with harness. (4) An explosion-proof lamp...

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

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

Williams, M. G.; Mouser, M. R.; Simon, J. B.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditionsmore » in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed to be reliable in these conditions. The primary goal of any such actions is to maintain or refill the passive inventory available to cool the core, containment and spent fuel pool in the safety-related and seismically qualified Passive Containment Cooling Water Storage Tank (PCCWST). The seismically-qualified, ground-mounted Passive Containment Cooling Ancillary Water Storage Tank (PCCAWST) is also available for this function as appropriate. The primary effect of these actions would be to increase the coping time for the AP1000 during design basis events, as well as events such as those described above, from 72 hours without operator intervention to 7 days with minimal operator actions. These Operator actions necessary to protect the health and safety of the public are addressed in the Post-72 Hour procedures, as well as some EOPs, AOPs, ARPs and the Severe Accident Management Guidelines (SAMGs). Should the event continue to become more severe and plant conditions degrade further with indications of inadequate core cooling, the SAMGs provide guidance for strategies to address these hypothetical severe accident conditions. The AP1000 SAMG diagnoses and actions are prioritized to first utilize the AP1000 features that are expected to retain a damaged core inside the reactor vessel. Only one strategy is undertaken at any time. This strategy will be followed and its effectiveness evaluated before other strategies are undertaken. This is a key feature of both the symptom-oriented AP1000 EOPs and the AP1000 SAMGs which maximizes the probability of retaining a damaged core inside the reactor vessel and containment while minimizing the chances for confusion and human errors during implementation. The AP1000 SAMGs are simple and straight-forward and have been developed with considerable input from human factors and plant operations experts. Most importantly, and different from severe accident management strategies for other plants, the AP1000 SAMGs do not require diagnosis of the location of the core (i.e., whether reactor vessel failure has occurred). This is a fundamental consequence of the AP1000 In-Vessel Retention approach, which allows severe accident management to be based on fundamental principles (e.g. provide coolant as close as possible to the core) that do not change during a specific event. This eliminates the need for one of the more difficult diagnostic requirements, since reactor vessel failure does not directly relate to any measurable plant parameter, and differs from other designs in that an engineered failure of the pressure vessel' (e.g. core catcher) is never required. (authors)« less

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

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.

[The relevance of core muscles in ice hockey players: a feasibility study].

PubMed

Rogan, S; Blasimann, A; Nyffenegger, D; Zimmerli, N; Radlinger, L

2013-12-01

Good core strength is seen as a condition for high performance in sports. In general, especially maximum voluntary contraction (MVC) and strength endurance (SE) measurements of the core muscles are used. In addition, a few studies can be found that examine the core muscles in terms of MVC, rate of force development (RFD) and SE. Primary aims of this feasibility study were to investigate the feasibility regarding recruiting process, compliance and safety of the testing conditions and raise the force capabilities MVC, RFD and SE of the core muscles in amateur ice hockey players. Secondarily, tendencies of correlations between muscle activity and either shot speed and sprint time shall be examined. In this feasibility study the recruitment process has been approved by 29 ice hockey players, their adherence to the study measurements of trunk muscles, and safety of the measurements was evaluated. To determine the MVC, RFD and SE for the ventral, lateral and dorsal core muscles a dynamic force measurement was performed. To determine the correlation between core muscles and shot speed and 40-m sprint, respectively, the rank correlation coefficient (rho) from Spearman was used. The recruited number of eight field players and one goal-keeper was not very high. The compliance with 100 % was excellent. The players reported no adverse symptoms or injuries after the measurements. The results show median values for the ventral core muscles for MVC with 46.5 kg for RFD with 2.23 m/s2 and 96 s for the SE. For lateral core muscle median values of the lateral core muscles for MVC with 71.10 kg, RFD with 2.59 m/s2 and for SE over 66 s were determined. The dorsal core muscles shows values for MVC 69.7 kg, for RFD 3.39 m/s2 and for SE of 75 s. High correlations between MVC of the ventral core muscles (rho = -0.721, p = 0.021), and between the SE of the ventral core muscles (rho = 0.787, p = 0.012), and the shot velocity rate were determined. Another high correlation between SE of the ventral core muscles and sprint over 40 m (rho = 0.717, p = 0.030) could be demonstrated. This feasibility study has shown that the implementation of the selected design is adapted for future studies. Further studies are needed to better understand the relationship between the velocity rate and the MVC, and the SE respectively, as well as between the sprint and the SE. © Georg Thieme Verlag KG Stuttgart · New York.

An assessment and validation study of nuclear reactors for low power space applications

NASA Technical Reports Server (NTRS)

Klein, A. C.; Gedeon, S. R.; Morey, D. C.

1987-01-01

The feasibility and safety of six conceptual small, low power nuclear reactor designs was evaluated. Feasibility evaluations included the determination of sufficient reactivity margins for seven years of full power operation and safe shutdown as well as handling during pre-launch assembly phases. Safety evaluations were concerned with the potential for maintaining subcritical conditions in the event of launch or transportation accidents. These included water immersion accident scenarios both with and without water flooding the core. Results show that most of the concepts can potentially meet the feasibility and safety requirements; however, due to the preliminary nature of the designs considered, more detailed designs will be necessary to enable these concepts to fully meet the safety requirements.

Tank 241-AZ-102 Privatization Push Mode Core Sampling and Analysis Plan

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

RASMUSSEN, J.H.

1999-08-02

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AZ-102. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AZ-102 required to satisfy the Data Quality Objectives For TWRS Privatization Phase I: Confirm Tank TIS An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase 1: Confirm Tank TIS An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste andmore » High Level Waste Feed Data Quality Objectives (L&H DQO) (Patello et al. 1999) and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). The Tank Characterization Technical Sampling Basis document (Brown et al. 1998) indicates that these issues, except the Equipment DQO apply to tank 241-AZ-102 for this sampling event. The Equipment DQO is applied for shear strength measurements of the solids segments only. Poppiti (1999) requires additional americium-241 analyses of the sludge segments. Brown et al. (1998) also identify safety screening, regulatory issues and provision of samples to the Privatization Contractor(s) as applicable issues for this tank. However, these issues will not be addressed via this sampling event. Reynolds et al. (1999) concluded that information from previous sampling events was sufficient to satisfy the safety screening requirements for tank 241 -AZ-102. Push mode core samples will be obtained from risers 15C and 24A to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples, composite the liquids and solids, perform chemical analyses, and provide subsamples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AZ-102 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plan.« less

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

Design criteria for a self-actuated shutdown system to ensure limitation of core damage. [LMFBR

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

Deane, N.A.; Atcheson, D.B.

1981-09-01

Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times.

Measuring Best Practices for Workplace Safety, Health, and Well-Being: The Workplace Integrated Safety and Health Assessment.

PubMed

Sorensen, Glorian; Sparer, Emily; Williams, Jessica A R; Gundersen, Daniel; Boden, Leslie I; Dennerlein, Jack T; Hashimoto, Dean; Katz, Jeffrey N; McLellan, Deborah L; Okechukwu, Cassandra A; Pronk, Nicolaas P; Revette, Anna; Wagner, Gregory R

2018-05-01

To present a measure of effective workplace organizational policies, programs, and practices that focuses on working conditions and organizational facilitators of worker safety, health and well-being: the workplace integrated safety and health (WISH) assessment. Development of this assessment used an iterative process involving a modified Delphi method, extensive literature reviews, and systematic cognitive testing. The assessment measures six core constructs identified as central to best practices for protecting and promoting worker safety, health and well-being: leadership commitment; participation; policies, programs, and practices that foster supportive working conditions; comprehensive and collaborative strategies; adherence to federal and state regulations and ethical norms; and data-driven change. The WISH Assessment holds promise as a tool that may inform organizational priority setting and guide research around causal pathways influencing implementation and outcomes related to these approaches.

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2013 CFR

2013-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2014 CFR

2014-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

29 CFR 1926.104 - Safety belts, lifelines, and lanyards.

Code of Federal Regulations, 2010 CFR

2010-07-01

... be subjected to cutting or abrasion, shall be a minimum of 7/8-inch wire core manila rope. For all... or pressed steel, cadmium plated in accordance with type 1, Class B plating specified in Federal...

29 CFR 1926.104 - Safety belts, lifelines, and lanyards.

Code of Federal Regulations, 2011 CFR

2011-07-01

... be subjected to cutting or abrasion, shall be a minimum of 7/8-inch wire core manila rope. For all... or pressed steel, cadmium plated in accordance with type 1, Class B plating specified in Federal...

Safety pharmacology--current and emerging concepts.

PubMed

Hamdam, Junnat; Sethu, Swaminathan; Smith, Trevor; Alfirevic, Ana; Alhaidari, Mohammad; Atkinson, Jeffrey; Ayala, Mimieveshiofuo; Box, Helen; Cross, Michael; Delaunois, Annie; Dermody, Ailsa; Govindappa, Karthik; Guillon, Jean-Michel; Jenkins, Rosalind; Kenna, Gerry; Lemmer, Björn; Meecham, Ken; Olayanju, Adedamola; Pestel, Sabine; Rothfuss, Andreas; Sidaway, James; Sison-Young, Rowena; Smith, Emma; Stebbings, Richard; Tingle, Yulia; Valentin, Jean-Pierre; Williams, Awel; Williams, Dominic; Park, Kevin; Goldring, Christopher

2013-12-01

Safety pharmacology (SP) is an essential part of the drug development process that aims to identify and predict adverse effects prior to clinical trials. SP studies are described in the International Conference on Harmonisation (ICH) S7A and S7B guidelines. The core battery and supplemental SP studies evaluate effects of a new chemical entity (NCE) at both anticipated therapeutic and supra-therapeutic exposures on major organ systems, including cardiovascular, central nervous, respiratory, renal and gastrointestinal. This review outlines the current practices and emerging concepts in SP studies including frontloading, parallel assessment of core battery studies, use of non-standard species, biomarkers, and combining toxicology and SP assessments. Integration of the newer approaches to routine SP studies may significantly enhance the scope of SP by refining and providing mechanistic insight to potential adverse effects associated with test compounds. Copyright © 2013 Elsevier Inc. All rights reserved.

Nuclear reactor safety device

DOEpatents

Hutter, E.

1983-08-15

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

The importance of data collection for timely and accurate risk assessment

NASA Astrophysics Data System (ADS)

Gilsenan, MB

2017-09-01

The European Food Safety Authority (EFSA) is responsible for food safety risk assessments at EU level. It provides independent scientific advice on risks associated with the food chain to support EU risk management decisions. Since its establishment, EFSA has amassed a wealth of data to underpin its risk assessments, such as food consumption data, monitoring data and experimental data. Increasing transparency of its risk assessments is a core objective of EFSA. EFSA aims to enhance the quality and transparency of its outputs by giving insofar as possible access to data and methods underpinning its scientific outputs. This paper provides an overview of the role of EFSA, its core data collections and their regulatory framework, as well as data quality and standardisation aspects. Finally, the paper elaborates on EFSA’s 2020 strategy in relation to data, and describes EFSA scientific data warehouse and Knowledge Junction in this regard.

An Interoperability Platform Enabling Reuse of Electronic Health Records for Signal Verification Studies

PubMed Central

Yuksel, Mustafa; Gonul, Suat; Laleci Erturkmen, Gokce Banu; Sinaci, Ali Anil; Invernizzi, Paolo; Facchinetti, Sara; Migliavacca, Andrea; Bergvall, Tomas; Depraetere, Kristof; De Roo, Jos

2016-01-01

Depending mostly on voluntarily sent spontaneous reports, pharmacovigilance studies are hampered by low quantity and quality of patient data. Our objective is to improve postmarket safety studies by enabling safety analysts to seamlessly access a wide range of EHR sources for collecting deidentified medical data sets of selected patient populations and tracing the reported incidents back to original EHRs. We have developed an ontological framework where EHR sources and target clinical research systems can continue using their own local data models, interfaces, and terminology systems, while structural interoperability and Semantic Interoperability are handled through rule-based reasoning on formal representations of different models and terminology systems maintained in the SALUS Semantic Resource Set. SALUS Common Information Model at the core of this set acts as the common mediator. We demonstrate the capabilities of our framework through one of the SALUS safety analysis tools, namely, the Case Series Characterization Tool, which have been deployed on top of regional EHR Data Warehouse of the Lombardy Region containing about 1 billion records from 16 million patients and validated by several pharmacovigilance researchers with real-life cases. The results confirm significant improvements in signal detection and evaluation compared to traditional methods with the missing background information. PMID:27123451

A high-fidelity Monte Carlo evaluation of CANDU-6 safety parameters

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

Kim, Y.; Hartanto, D.

2012-07-01

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium) reactor have been evaluated by using a modified MCNPX code. For accurate analysis of the parameters, the DBRC (Doppler Broadening Rejection Correction) scheme was implemented in MCNPX in order to account for the thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted by using the MCNPX and the FTC value is evaluated for several burnup points including the mid-burnupmore » representing a near-equilibrium core. The Doppler effect has been evaluated by using several cross section libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR value is also evaluated at mid-burnup conditions to characterize safety features of equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, huge number of neutron histories are considered in this work and the standard deviation of the k-inf values is only 0.5{approx}1 pcm. It has been found that the FTC is significantly enhanced by accounting for the Doppler broadening of scattering resonance and the PCR are clearly improved. (authors)« less

On stress-state optimization in steel-concrete composite structures

NASA Astrophysics Data System (ADS)

Brauns, J.; Skadins, U.

2017-10-01

The plastic resistance of a concrete-filled column commonly is given as a sum of the components and taking into account the effect of confinement. The stress state in a composite column is determined by taking into account the non-linear relationship of modulus of elasticity and Poisson’s ratio on the stress level in the concrete core. The effect of confinement occurs at a high stress level when structural steel acts in tension and concrete in lateral compression. The stress state of a composite beam is determined taking into account non-linear dependence on the position of neutral axis. In order to improve the stress state of a composite element and increase the safety of the construction the appropriate strength of steel and concrete has to be applied. The safety of high-stressed composite structures can be achieved by using high-performance concrete (HPC). In this study stress analysis of the composite column and beam is performed with the purpose of obtaining the maximum load-bearing capacity and enhance the safety of the structure by using components with the appropriate strength and by taking into account the composite action. The effect of HPC on the stress state and load carrying capacity of composite elements is analysed.

Safety climate and the theory of planned behavior: towards the prediction of unsafe behavior.

PubMed

Fogarty, Gerard J; Shaw, Andrew

2010-09-01

The present study is concerned with the human factors that contribute to violations in aviation maintenance. Much of our previous research in this area has been based on safety climate surveys and the analysis of relations among core dimensions of climate. In this study, we tap into mainstream psychological theory to help clarify the mechanisms underlying the links between climate and behavior. Specifically, we demonstrate the usefulness of Ajzen's (1991, 2001) Theory of Planned Behavior (TPB) to understanding violation behaviors in aircraft maintenance. A questionnaire was administered to 307 aircraft maintenance workers. Constructs measured by the survey included perceptions of management attitudes to safety, own attitudes to violations, intention to violate, group norms, workplace pressures, and violations. A model based on the TPB illustrated hypothetical connections among these variables. Path analyses using AMOS suggested some theoretically justifiable modifications to the model. Fit statistics of the revised model were excellent with intentions, group norms, and personal attitudes combining to explain 50% of the variance in self-reported violations. The model highlighted the importance of management attitudes and group norms as direct and indirect predictors of violation behavior. We conclude that the TPB is a useful tool for understanding the psychological background to the procedural violations so often associated with incidents and accidents. 2009 Elsevier Ltd. All rights reserved.

A review for identification of initiating events in event tree development process on nuclear power plants

NASA Astrophysics Data System (ADS)

Riyadi, Eko H.

2014-09-01

Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.

Research on removing reservoir core water sensitivity using the method of ultrasound-chemical agent for enhanced oil recovery.

PubMed

Wang, Zhenjun; Huang, Jiehao

2018-04-01

The phenomenon of water sensitivity often occurs in the oil reservoir core during the process of crude oil production, which seriously affects the efficiency of oil extraction. In recent years, near-well ultrasonic processing technology attaches more attention due to its safety and energy efficient. In this paper, the comparison of removing core water sensitivity by ultrasonic wave, chemical injection and ultrasound-chemical combination technique are investigated through experiments. Results show that: lower ultrasonic frequency and higher power can improve the efficiency of core water sensitivity removal; the effects of removing core water sensitivity under ultrasonic treatment get better with increase of core initial permeability; the effect of removing core water sensitivity using ultrasonic treatment won't get better over time. Ultrasonic treatment time should be controlled in a reasonable range; the effect of removing core water sensitivity using chemical agent alone is slightly better than that using ultrasonic treatment, however, chemical injection could be replaced by ultrasonic treatment for removing core water sensitivity from the viewpoint of oil reservoir protection and the sustainable development of oil field; ultrasound-chemical combination technique has the best effect for water sensitivity removal than using ultrasonic treatment or chemical injection alone. Copyright © 2017 Elsevier B.V. All rights reserved.

Controlled-Release Microcapsules for Smart Coatings for Corrosion Applications

NASA Technical Reports Server (NTRS)

2008-01-01

Corrosion is a serious problem that has enormous costs and serious safety implications. Localized corrosion, such as pitting, is very dangerous and can cause catastrophic failures. The NASA Corrosion Technology Laboratory at Kennedy Space Center is developing a smart coating based on pH-sensitive microcapsules for corrosion applications. These versatile microcapsules are designed to be incorporated into a smart coating and deliver their core content when corrosion starts. Corrosion indication was the first function incorporated into the microcapsules. Current efforts are focused on incorporating the corrosion inhibition function through the encapsulation of corrosion inhibitors into water core and oil core microcapsules. Scanning electron microscopy (SEM) images of encapsulated corrosion inhibitors are shown.

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

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

Philip E. MacDonald

2005-01-01

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission ofmore » the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.« less

Leveraging the Continuum: A Novel Approach to Meeting Quality Improvement and Patient Safety Competency Requirements Across a Large Department of Medicine.

PubMed

Myers, Jennifer S; Bellini, Lisa M

2018-05-22

Quality improvement (QI) and patient safety (PS) are now core competencies across the medical education continuum. A major challenge to developing and implementing these new curricular requirements is the lack of faculty expertise. In 2015, the authors developed a centralized, vertically integrated, competency-based approach to meet the educational requirements in QI/PS across the continuum of graduate medical education in the Department of Medicine, Perelman School of Medicine, University of Pennsylvania. By leveraging the QI/PS expertise of one individual, the authors identified and trained core QI/PS faculty members and sequentially deployed QI/PS activities that were tailored to the learner level and specialty. The curriculum includes PS event reporting, systems thinking and root cause analysis skills, adverse event disclosure, and a QI workshop series and project. PS event reporting, an indication of engagement in PS culture, increased by 186% among interns, 384% among PGY 2 and 3 residents, and 613% among fellows between AYs 2013-2014 and 2016-2017. In AY 2017-2018, 9 faculty members and 40 fellows from 9 fellowships participated in the QI workshop series, and 53 fellows from 7 fellowships participated in the adverse event disclosure simulation activity. All educational activities were rated highly. The authors are expanding the adverse event disclosure activity to include residents and the remaining fellowship programs, identifying fellowships to pilot curricular efforts related to clinical quality metrics, developing introductory activities in basic QI/PS concepts for medical students, and evaluating the impact of efforts on the participating faculty members.

Integrating High-Dimensional Transcriptomics and Image Analysis Tools into Early Safety Screening: Proof of Concept for a New Early Drug Development Strategy.

PubMed

Verbist, Bie M P; Verheyen, Geert R; Vervoort, Liesbet; Crabbe, Marjolein; Beerens, Dominiek; Bosmans, Cindy; Jaensch, Steffen; Osselaer, Steven; Talloen, Willem; Van den Wyngaert, Ilse; Van Hecke, Geert; Wuyts, Dirk; Van Goethem, Freddy; Göhlmann, Hinrich W H

2015-10-19

During drug discovery and development, the early identification of adverse effects is expected to reduce costly late-stage failures of candidate drugs. As risk/safety assessment takes place rather late during the development process and due to the limited ability of animal models to predict the human situation, modern unbiased high-dimensional biology readouts are sought, such as molecular signatures predictive for in vivo response using high-throughput cell-based assays. In this theoretical proof of concept, we provide findings of an in-depth exploration of a single chemical core structure. Via transcriptional profiling, we identified a subset of close analogues that commonly downregulate multiple tubulin genes across cellular contexts, suggesting possible spindle poison effects. Confirmation via a qualified toxicity assay (in vitro micronucleus test) and the identification of a characteristic aggregate-formation phenotype via exploratory high-content imaging validated the initial findings. SAR analysis triggered the synthesis of a new set of compounds and allowed us to extend the series showing the genotoxic effect. We demonstrate the potential to flag toxicity issues by utilizing data from exploratory experiments that are typically generated for target evaluation purposes during early drug discovery. We share our thoughts on how this approach may be incorporated into drug development strategies.

Laser Safety: A Laser Alignment Practical Training Course

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

Woods, Michael; Edstrom, Steve; /SLAC

2011-01-26

SLAC National Accelerator Laboratory has developed a Laser Alignment Practical Training Course as one of its core laser safety classes. The course is taught to small groups of up to three students and takes 1-3 hours to complete. This practical course is not a substitute for site-specific On-the-Job Training; it does, however, provide a good introduction in core laser safety practices that can be broadly applied. Alignment and diagnostic tasks are performed with low power lasers. Students learn safe alignment and diagnostic techniques and how to avoid common mistakes that might lead to an accident. The class is taught bymore » laser supervisors, enabling them to assess the skill level of new laser personnel and determine the subsequent level of supervision needed. The course has six alignment tasks. For each task, discussion points are given for the instructor to review with the students. The optics setup includes different wavelength lasers, a beam expander, mirrors, irises, a periscope, a beam-splitting polarizer and a diffraction grating. Diagnostic tools include viewing cards, an IR viewer and a ccd camera. Laser eyewear is available to block some laser wavelengths in the setup.« less

Endobronchial Ultrasound-Guided Cautery-Assisted Transbronchial Forceps Biopsies: Safety and Sensitivity Relative to Transbronchial Needle Aspiration.

PubMed

Bramley, Kyle; Pisani, Margaret A; Murphy, Terrence E; Araujo, Katy L; Homer, Robert J; Puchalski, Jonathan T

2016-05-01

Endobronchial ultrasound (EBUS)-guided transbronchial needle aspiration (TBNA) is important in the evaluation of thoracic lymphadenopathy. Reliably providing excellent diagnostic yield for malignancy, its diagnosis of sarcoidosis is inconsistent. Furthermore, TBNA may not suffice when larger "core biopsy" samples of malignant tissue are required. The primary objective of this study was to determine if the sequential use of TBNA and a novel technique called cautery-assisted transbronchial forceps biopsy (ca-TBFB) was safe. Secondary outcomes included sensitivity and successful acquisition of tissue. The study prospectively enrolled 50 unselected patients undergoing convex-probe EBUS. All lymph nodes exceeding 1 cm were sequentially biopsied under EBUS guidance using TBNA and ca-TBFB. Safety and sensitivity were assessed at the nodal level for 111 nodes. Results of each technique were also reported for each patient. There were no significant adverse events. In nodes determined to be malignant, TBNA provided higher sensitivity (100%) than ca-TBFB (78%). However, among nodes with granulomatous inflammation, ca-TBFB exhibited higher sensitivity (90%) than TBNA (33%). On the one hand, for analysis based on patients rather than nodes, 6 of the 31 patients with malignancy would have been missed or understaged if the diagnosis were based on samples obtained by ca-TBFB. On the other hand, 3 of 8 patients with sarcoidosis would have been missed if analysis were based only on TBNA samples. In some patients, only ca-TBFB acquired sufficient tissue for the core samples needed in clinical trials of malignancy. The sequential use of TBNA and ca-TBFB appears to be safe. The larger samples obtained from ca-TBFB increased its sensitivity to detect granulomatous disease and provided adequate specimens for clinical trials of malignancy when specimens from needle biopsies were insufficient. For thoracic surgeons and advanced bronchoscopists, we advocate ca-TBFB as an alternative to TBNA in select clinical scenarios. Copyright © 2016 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

Influence of core design, production technique, and material selection on fracture behavior of yttria-stabilized tetragonal zirconia polycrystal fixed dental prostheses produced using different multilayer techniques: split-file, over-pressing, and manually built-up veneers.

PubMed

Mahmood, Deyar Jallal Hadi; Linderoth, Ewa H; Wennerberg, Ann; Vult Von Steyern, Per

2016-01-01

To investigate and compare the fracture strength and fracture mode in eleven groups of currently, the most commonly used multilayer three-unit all-ceramic yttria-stabilized tetragonal zirconia polycrystal (Y-TZP) fixed dental prostheses (FDPs) with respect to the choice of core material, veneering material area, manufacturing technique, design of connectors, and radii of curvature of FDP cores. A total of 110 three-unit Y-TZP FDP cores with one intermediate pontic were made. The FDP cores in groups 1-7 were made with a split-file design, veneered with manually built-up porcelain, computer-aided design-on veneers, and over-pressed veneers. Groups 8-11 consisted of FDPs with a state-of-the-art design, veneered with manually built-up porcelain. All the FDP cores were subjected to simulated aging and finally loaded to fracture. There was a significant difference (P<0.05) between the core designs, but not between the different types of Y-TZP materials. The split-file designs with VITABLOCS(®) (1,806±165 N) and e.max(®) ZirPress (1,854±115 N) and the state-of-the-art design with VITA VM(®) 9 (1,849±150 N) demonstrated the highest mean fracture values. The shape of a split-file designed all-ceramic reconstruction calls for a different dimension protocol, compared to traditionally shaped ones, as the split-file design leads to sharp approximal indentations acting as fractural impressions, thus decreasing the overall strength. The design of a framework is a crucial factor for the load bearing capacity of an all-ceramic FDP. The state-of-the-art design is preferable since the split-file designed cores call for a cross-sectional connector area at least 42% larger, to have the same load bearing capacity as the state-of-the-art designed cores. All veneering materials and techniques tested in the study, split-file, over-press, built-up porcelains, and glass-ceramics are, with a great safety margin, sufficient for clinical use both anteriorly and posteriorly. Analysis of the fracture pattern shows differences between the milled veneers and over-pressed or built-up veneers, where the milled ones show numerically more veneer cracks and the other groups only show complete connector fractures.

LOFT L2-3 blowdown experiment safety analyses D, E, and G; LOCA analyses H, K, K1

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

Perryman, J.L.; Keeler, C.D.; Saukkoriipi, L.O.

1978-12-01

Three calculations using conservative off-nominal conditions and evaluation model options were made using RELAP4/MOD5 for blowdown-refill and RELAP4/MOD6 for reflood for Loss-of-Fluid Test Experiment L2-3 to support the experiment safety analysis effort. The three analyses are as follows: Analysis D: Loss of commercial power during Experiment L2-3; Analysis E: Hot leg quick-opening blowdown valve (QOBV) does not open during Experiment L2-3; and Analysis G: Cold leg QOBV does not open during Experiment L2-3. In addition, the results of three LOFT loss-of-coolant accident (LOCA) analyses using a power of 56.1 MW and a primary coolant system flow rate of 3.6 millionmore » 1bm/hr are presented: Analysis H: Intact loop 200% hot leg break; emergency core cooling (ECC) system B unavailable; Analysis K: Pressurizer relief valve stuck in open position; ECC system B unavailable; and Analysis K1: Same as analysis K, but using a primary coolant system flow rate of 1.92 million 1bm/hr (L2-4 pre-LOCE flow rate). For analysis D, the maximum cladding temperature reached was 1762/sup 0/F, 22 sec into reflood. In analyses E and G, the blowdowns were slower due to one of the QOBVs not functioning. The maximum cladding temperature reached in analysis E was 1700/sup 0/F, 64.7 sec into reflood; for analysis G, it was 1300/sup 0/F at the start of reflood. For analysis H, the maximum cladding temperature reached was 1825/sup 0/F, 0.01 sec into reflood. Analysis K was a very slow blowdown, and the cladding temperatures followed the saturation temperature of the system. The results of analysis K1 was nearly identical to analysis K; system depressurization was not affected by the primary coolant system flow rate.« less

Potential connected vehicle applications to enhance mobility, safety, and environmental security.

DOT National Transportation Integrated Search

2012-02-01

The connected vehicle research initiative is the core of the U.S. Department of Transportations intelligent : transportation system research program. The initiative is beginning to gain momentum in the research : community because of the developme...

Inspection of the Math Model Tools for On-Orbit Assessment of Impact Damage Report. Version 1.0

NASA Technical Reports Server (NTRS)

Harris, Charles E.; Raju, Ivatury S.; Piascik, Robert S.; Kramer White, Julie; Labbe, Steve G.; Rotter, Hank A.

2005-01-01

In Spring of 2005, the NASA Engineering Safety Center (NESC) was engaged by the Space Shuttle Program (SSP) to peer review the suite of analytical tools being developed to support the determination of impact and damage tolerance of the Orbiter Thermal Protection Systems (TPS). The NESC formed an independent review team with the core disciplines of materials, flight sciences, structures, mechanical analysis and thermal analysis. The Math Model Tools reviewed included damage prediction and stress analysis, aeroheating analysis, and thermal analysis tools. Some tools are physics-based and other tools are empirically-derived. Each tool was created for a specific use and timeframe, including certification, real-time pre-launch assessments, and real-time on-orbit assessments. The tools are used together in an integrated strategy for assessing the ramifications of impact damage to tile and RCC. The NESC teams conducted a peer review of the engineering data package for each Math Model Tool. This report contains the summary of the team observations and recommendations from these reviews.

Design and Evaluation of Glass/epoxy Composite Blade and Composite Tower Applied to Wind Turbine

NASA Astrophysics Data System (ADS)

Park, Hyunbum

2018-02-01

In the study, the analysis and manufacturing of small class wind turbine blade was performed. In the structural design, firstly the loading conditions are defined through the load case analysis. The proposed structural configuration of blade has a sandwich type composite structure with the E-glass/Epoxy face sheets and the Urethane foam core for lightness, structural stability, low manufacturing cost and easy manufacturing process. And also, this work proposes a design procedure and results of tower for the small scale wind turbine systems. Structural analysis of blade including load cases, stress, deformation, buckling, vibration and fatigue life was performed using the finite element method, the load spectrum analysis and the Miner rule. Moreover, investigation on structural safety of tower was verified through structural analysis by FEM. The manufacturing of blade and tower was performed based on structural design. In order to investigate the designed structure, the structural tests were conducted and its results were compared with the calculated results. It is confirmed that the final proposed blade and tower meet the design requirements.

Accuracy and safety of ultrasound-guided percutaneous needle core biopsy of renal masses

PubMed Central

Wang, Xianding; Lv, Yuanhang; Xu, Zilin; Aniu, Muguo; Qiu, Yang; Wei, Bing; Li, Xiaohong; Wei, Qiang; Dong, Qiang; Lin, Tao

2018-01-01

Abstract Our aim is to determine the sufficiency, accuracy, and safety of ultrasound-guided percutaneous needle core biopsy of renal masses in Chinese patients. Patients who had undergone ultrasound-guided needle core renal mass biopsy from June 2012 to June 2016 at West China Hospital, China were retrospectively reviewed. The information obtained included demographics, mass-related parameters, biopsy indications, technique, complications, pathologic results, and follow-up. Concordance of surgical resection pathology and follow-up data were assessed. Renal mass biopsies were performed in 106 patients. Thirty-nine (36.8%) were asymptomatic. The male/female ratio was 60/46, with a median age of 49.5 years. Median mass size was 8.1 cm (range 1.8–20). Biopsy was performed through a 16-gauge needle, with median cores of 2 taken (range 1–5). Only one significant biopsy-related complication (hemorrhage requiring transfusion) was encountered. An adequate tissue sample was obtained in 97.2% (103/106) of biopsies. Eighty-seven biopsies (82.1%) showed malignant neoplasms, 16 (15.1%) yielded benignity, and 3 (2.8%) were nondiagnostic. After biopsy, 46 patients (43.4%) underwent surgery. Compared with the subsequent mass resection pathology, the biopsy diagnoses were identical in 43 cases. The accuracy rate of biopsy distinguishing malignant from benign lesions was 99.1%, and the rate for determining tumor histological type (excluding the nondiagnostic biopsies) was 95.1%. The sensitivity and specificity in detecting malignancy were 98.9% and 100%, respectively. In several situations, there is still a role for biopsy before intervention. Percutaneous needle core biopsy under ultrasonography guidance is highly accurate and safe, and can determine the proper management of undefinable masses. PMID:29595650

Biocompatibility and safety of a hybrid core-shell nanoparticulate OP-1 delivery system intramuscularly administered in rats.

PubMed

Haidar, Ziyad S; Hamdy, Reggie C; Tabrizian, Maryam

2010-04-01

A hybrid, localized and release-controlled delivery system for bone growth factors consisting of a liposomal core incorporated into a shell of alternating layer-by-layer self-assembled natural polyelectrolytes has been formulated. Hydrophilic, monodisperse, spherical and stable cationic nanoparticles (< or =350 nm) with an extended shelf-life resulted. Cytocompatibility was previously assayed with MC3T3-E1.4 mouse preosteoblasts showing no adverse effects on cell viability. In this study, the in vivo biocompatibility of unloaded and loaded nanoparticles with osteogenic protein-1 or OP-1 was investigated. Young male Wistar rats were injected intramuscularly and monitored over a period of 10 weeks for signs of inflammation and/or adverse reactions. Blood samples (600 microL/collection) were withdrawn followed by hematological and biochemical analysis. Body weight changes over the treatment period were noted. Major organs were harvested, weighed and examined histologically for any pathological changes. Finally, the injection site was identified and examined immunohistochemically. Overall, all animals showed no obvious toxic health effects, immune responses and/or change in organ functions. This hybrid core-shell nanoparticulate delivery system localizes the effect of the released bioactive load within the site of injection in muscle with no significant tissue distress. Hence, a safe and promising carrier for therapeutic growth factors and possibly other biomolecules is presented. 2009 Elsevier Ltd. All rights reserved.

Hedyotis diffusa Combined with Scutellaria barbata Are the Core Treatment of Chinese Herbal Medicine Used for Breast Cancer Patients: A Population-Based Study

PubMed Central

Yeh, Yuan-Chieh; Chen, Hsing-Yu; Yang, Sien-Hung; Lin, Yi-Hsien; Chiu, Jen-Hwey; Lin, Yi-Hsuan; Chen, Jiun-Liang

2014-01-01

Traditional Chinese medicine (TCM), which is the most common type of complementary and alternative medicine (CAM) used in Taiwan, is increasingly used to treat patients with breast cancer. However, large-scale studies on the patterns of TCM prescriptions for breast cancer are still lacking. The aim of this study was to determine the core treatment of TCM prescriptions used for breast cancer recorded in the Taiwan National Health Insurance Research Database. TCM visits made for breast cancer in 2008 were identified using ICD-9 codes. The prescriptions obtained at these TCM visits were evaluated using association rule mining to evaluate the combinations of Chinese herbal medicine (CHM) used to treat breast cancer patients. A total of 37,176 prescriptions were made for 4,436 outpatients with breast cancer. Association rule mining and network analysis identified Hedyotis diffusa plus Scutellaria barbata as the most common duplex medicinal (10.9%) used for the core treatment of breast cancer. Jia-Wei-Xiao-Yao-San (19.6%) and Hedyotis diffusa (41.9%) were the most commonly prescribed herbal formula (HF) and single herb (SH), respectively. Only 35% of the commonly used CHM had been studied for efficacy. More clinical trials are needed to evaluate the efficacy and safety of these CHM used to treat breast cancer. PMID:24734104

Hedyotis diffusa Combined with Scutellaria barbata Are the Core Treatment of Chinese Herbal Medicine Used for Breast Cancer Patients: A Population-Based Study.

PubMed

Yeh, Yuan-Chieh; Chen, Hsing-Yu; Yang, Sien-Hung; Lin, Yi-Hsien; Chiu, Jen-Hwey; Lin, Yi-Hsuan; Chen, Jiun-Liang

2014-01-01

Traditional Chinese medicine (TCM), which is the most common type of complementary and alternative medicine (CAM) used in Taiwan, is increasingly used to treat patients with breast cancer. However, large-scale studies on the patterns of TCM prescriptions for breast cancer are still lacking. The aim of this study was to determine the core treatment of TCM prescriptions used for breast cancer recorded in the Taiwan National Health Insurance Research Database. TCM visits made for breast cancer in 2008 were identified using ICD-9 codes. The prescriptions obtained at these TCM visits were evaluated using association rule mining to evaluate the combinations of Chinese herbal medicine (CHM) used to treat breast cancer patients. A total of 37,176 prescriptions were made for 4,436 outpatients with breast cancer. Association rule mining and network analysis identified Hedyotis diffusa plus Scutellaria barbata as the most common duplex medicinal (10.9%) used for the core treatment of breast cancer. Jia-Wei-Xiao-Yao-San (19.6%) and Hedyotis diffusa (41.9%) were the most commonly prescribed herbal formula (HF) and single herb (SH), respectively. Only 35% of the commonly used CHM had been studied for efficacy. More clinical trials are needed to evaluate the efficacy and safety of these CHM used to treat breast cancer.

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

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

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

NASA Astrophysics Data System (ADS)

Khuwaileh, Bassam

High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL) based algorithm previously developed to quantify the uncertainty for single physics models is extended for large scale multi-physics coupled problems with feedback effect. Moreover, a non-linear surrogate based UQ approach is developed, used and compared to performance of the KL approach and brute force Monte Carlo (MC) approach. On the other hand, an efficient Data Assimilation (DA) algorithm is developed to assess information about model's parameters: nuclear data cross-sections and thermal-hydraulics parameters. Two improvements are introduced in order to perform DA on the high dimensional problems. First, a goal-oriented surrogate model can be used to replace the original models in the depletion sequence (MPACT -- COBRA-TF - ORIGEN). Second, approximating the complex and high dimensional solution space with a lower dimensional subspace makes the sampling process necessary for DA possible for high dimensional problems. Moreover, safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. Accordingly, an inverse problem can be defined and solved to assess the contributions from sources of uncertainty; and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this dissertation a subspace-based gradient-free and nonlinear algorithm for inverse uncertainty quantification namely the Target Accuracy Assessment (TAA) has been developed and tested. The ideas proposed in this dissertation were first validated using lattice physics applications simulated using SCALE6.1 package (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lattice models). Ultimately, the algorithms proposed her were applied to perform UQ and DA for assembly level (CASL progression problem number 6) and core wide problems representing Watts Bar Nuclear 1 (WBN1) for cycle 1 of depletion (CASL Progression Problem Number 9) modeled via simulated using VERA-CS which consists of several multi-physics coupled models. The analysis and algorithms developed in this dissertation were encoded and implemented in a newly developed tool kit algorithms for Reduced Order Modeling based Uncertainty/Sensitivity Estimator (ROMUSE).

Thinking in Pharmacy Practice: A Study of Community Pharmacists’ Clinical Reasoning in Medication Supply Using the Think-Aloud Method

PubMed Central

Croft, Hayley; Gilligan, Conor; Rasiah, Rohan; Levett-Jones, Tracy; Schneider, Jennifer

2017-01-01

Medication review and supply by pharmacists involves both cognitive and technical skills related to the safety and appropriateness of prescribed medicines. The cognitive ability of pharmacists to recall, synthesise and memorise information is a critical aspect of safe and optimal medicines use, yet few studies have investigated the clinical reasoning and decision-making processes pharmacists use when supplying prescribed medicines. The objective of this study was to examine the patterns and processes of pharmacists’ clinical reasoning and to identify the information sources used, when making decisions about the safety and appropriateness of prescribed medicines. Ten community pharmacists participated in a simulation in which they were required to review a prescription and make decisions about the safety and appropriateness of supplying the prescribed medicines to the patient, whilst at the same time thinking aloud about the tasks required. Following the simulation each pharmacist was asked a series of questions to prompt retrospective thinking aloud using video-stimulated recall. The simulated consultation and retrospective interview were recorded and transcribed for thematic analysis. All of the pharmacists made a safe and appropriate supply of two prescribed medicines to the simulated patient. Qualitative analysis identified seven core thinking processes used during the supply process: considering prescription in context, retrieving information, identifying medication-related issues, processing information, collaborative planning, decision making and reflection; and align closely with other health professionals. The insights from this study have implications for enhancing awareness of decision making processes in pharmacy practice and informing teaching and assessment approaches in medication supply. PMID:29301223

Thinking in Pharmacy Practice: A Study of Community Pharmacists' Clinical Reasoning in Medication Supply Using the Think-Aloud Method.

PubMed

Croft, Hayley; Gilligan, Conor; Rasiah, Rohan; Levett-Jones, Tracy; Schneider, Jennifer

2017-12-31

Medication review and supply by pharmacists involves both cognitive and technical skills related to the safety and appropriateness of prescribed medicines. The cognitive ability of pharmacists to recall, synthesise and memorise information is a critical aspect of safe and optimal medicines use, yet few studies have investigated the clinical reasoning and decision-making processes pharmacists use when supplying prescribed medicines. The objective of this study was to examine the patterns and processes of pharmacists' clinical reasoning and to identify the information sources used, when making decisions about the safety and appropriateness of prescribed medicines. Ten community pharmacists participated in a simulation in which they were required to review a prescription and make decisions about the safety and appropriateness of supplying the prescribed medicines to the patient, whilst at the same time thinking aloud about the tasks required. Following the simulation each pharmacist was asked a series of questions to prompt retrospective thinking aloud using video-stimulated recall. The simulated consultation and retrospective interview were recorded and transcribed for thematic analysis. All of the pharmacists made a safe and appropriate supply of two prescribed medicines to the simulated patient. Qualitative analysis identified seven core thinking processes used during the supply process: considering prescription in context, retrieving information, identifying medication-related issues, processing information, collaborative planning, decision making and reflection; and align closely with other health professionals. The insights from this study have implications for enhancing awareness of decision making processes in pharmacy practice and informing teaching and assessment approaches in medication supply.

Biophysical aspects of human thermoregulation during heat stress.

PubMed

Cramer, Matthew N; Jay, Ollie

2016-04-01

Humans maintain a relatively constant core temperature through the dynamic balance between endogenous heat production and heat dissipation to the surrounding environment. In response to metabolic or environmental disturbances to heat balance, the autonomic nervous system initiates cutaneous vasodilation and eccrine sweating to facilitate higher rates of dry (primarily convection and radiation) and evaporative transfer from the body surface; however, absolute heat losses are ultimately governed by the properties of the skin and the environment. Over the duration of a heat exposure, the cumulative imbalance between heat production and heat dissipation leads to body heat storage, but the consequent change in core temperature, which has implications for health and safety in occupational and athletic settings particularly among certain clinical populations, involves a complex interaction between changes in body heat content and the body's morphological characteristics (mass, surface area, and tissue composition) that collectively determine the body's thermal inertia. The aim of this review is to highlight the biophysical aspects of human core temperature regulation by outlining the principles of human energy exchange and examining the influence of body morphology during exercise and environmental heat stress. An understanding of the biophysical factors influencing core temperature will enable researchers and practitioners to better identify and treat individuals/populations most vulnerable to heat illness and injury during exercise and extreme heat events. Further, appropriate guidelines may be developed to optimize health, safety, and work performance during heat stress. Copyright © 2016 Elsevier B.V. All rights reserved.

Program Director Perceptions of Proficiency in the Core Entrustable Professional Activities.

PubMed

Pearlman, R Ellen; Pawelczak, Melissa; Yacht, Andrew C; Akbar, Salaahuddin; Farina, Gino A

2017-10-01

The Association of American Medical Colleges describes 13 core entrustable professional activities (EPAs) that every graduating medical student should be expected to perform proficiently on day 1 of residency, regardless of chosen specialty. Studies have shown wide variability in program director (PD) confidence in interns' abilities to perform these core EPAs. Little is known regarding comparison of United States Medical Licensing Examination (USMLE) scores with proficiency in EPAs. We determined if PDs from a large health system felt confident in their postgraduate year 1 residents' abilities to perform the 13 core EPAs, and compared perceived EPA proficiency with USMLE Step 1 and Step 2 scores. The PDs were asked to rate their residents' proficiency in each EPA and to provide residents' USMLE scores. Timing coincided with the reporting period for resident milestones. Surveys were completed on 204 of 328 residents (62%). PDs reported that 69% of residents (140 of 204) were prepared for EPA 4 (orders/prescriptions), 61% (117 of 192) for EPA 7 (form clinical questions), 68% (135 of 198) for EPA 8 (handovers), 63% (116 of 185) for EPA 11 (consent), and 38% (49 of 129) for EPA 13 (patient safety). EPA ratings and USMLE 1 and 2 were negatively correlated ( r (101) = -0.23, P  = .031). PDs felt that a significant percentage of residents were not adequately prepared in order writing, forming clinical questions, handoffs, informed consent, and promoting a culture of patient safety. We found no positive association between USMLE scores and EPA ratings.

Lithological properties of sedimentary environments in the shallow subsurface of the Northern Netherlands

NASA Astrophysics Data System (ADS)

Harting, Ronald; Bosch, Aleid; Gunnink, Jan

2014-05-01

Society has an increasing demand from the subsurface, which in the Dutch shallow subsurface (upper 30 to 40 meters) mainly focuses on natural aggregate resources, groundwater, infrastructure and dike safety. This stimulates the demand for knowledge about the composition and heterogeneity of the subsurface and its physical and chemical properties, including the uncertainties involved. Physical and chemical properties of sediments in the subsurface have been under investigation for decades; however, the usefulness of this data for applied research and the understanding of these properties is limited. This is due to several factors: studies consist mainly of separately collected datasets, targeted at a limited amount of parameters, focused on a small number of geological units, distributed unevenly with depth and usually collected from clustered drillings with limited spatial extent or are analysed with different techniques and methods, often on disturbed samples. These factors result in a heterogeneous and biased dataset not suitable to function as a reference dataset or to statistically determine regional characteristics of geological units. To overcome these shortcomings, the Geological Survey of the Netherlands is establishing a nation-wide reference dataset for physical and chemical properties. In 2006, a drilling campaign was started using cone penetration tests, cored drillings and geophysical well logs, choosing the sites for a good geographical distribution. The lithological properties of the undisturbed cores are visually described and interpreted for lithostratigraphy and inferred sedimentary environment based on lithofacies. The location of the samples in the cores are chosen based on this description and interpretation, resulting in an evenly distributed dataset of in situ samples with respect to geological units as well as an adequate number of samples suitable for statistical analysis. Analyses are uniformly performed for grain size distribution, permeability (both high and low permeable lithologies) and geochemical methods (X-Ray Fluorescence, Thermo-Gravimetric Analysis, Total Carbon, Total Sulphur and Total Organic Carbon). These analyses result in a large number of lithological, hydrological and geochemical parameters, i.e. clay content, sand median, vertical and horizontal permeability and CaCO3-content. We present the results from the analysis of lithological properties for the Northern Netherlands. Besides geology, these properties can be applied directly in studies concerning (amongst others) groundwater, natural aggregates and dike safety. We demonstrate the use of sedimentary environments based on lithofacies as a useful tool for comparison between lithostratigraphic units and lithofacies. These lithofacies match distinct parts of the marine, fluvial, glacial, eolian or organogenic environment, i.e. tidal channel sand, floodbasin clay and subglacial till. This results in lithological properties illustrating the heterogeneity within a geological unit and between equal depositional environments in different lithostratigraphic units. The acquired data have so far been used in several applied studies, i.e. improving parameterisation of 3D models leading to increased accuracy in groundwater models and dike safety studies concerning dike failure due to undermining. Recently, grain size distributions measured with different methods were recalibrated into a homogeneous dataset using this reference set, which greatly enlarged the dataset to be incorporated in the parameterisation of a 3D voxel model.

Patient safety: the landscape of the global research output and gender distribution.

PubMed

Schreiber, Moritz; Klingelhöfer, Doris; Groneberg, David A; Brüggmann, Doerthe

2016-02-12

Patient safety is a crucial issue in medicine. Its main objective is to reduce the number of deaths and health damages that are caused by preventable medical errors. To achieve this, it needs better health systems that make mistakes less likely and their effects less detrimental without blaming health workers for failures. Until now, there is no in-depth scientometric analysis on this issue that encompasses the interval between 1963 and 2014. Therefore, the aim of this study is to sketch a landscape of the past global research output on patient safety including the gender distribution of the medical discipline of patient safety by interpreting scientometric parameters. Additionally, respective future trends are to be outlined. The Core Collection of the scientific database Web of Science was searched for publications with the search term 'Patient Safety' as title word that was focused on the corresponding medical discipline. The resulting data set was analysed by using the methodology implemented by the platform NewQIS. To visualise the geographical landscape, state-of-the-art techniques including density-equalising map projections were applied. 4079 articles on patient safety were identified in the period from 1900 to 2014. Most articles were published in North America, the UK and Australia. In regard to the overall number of publications, the USA is the leading country, while the output ratio to the population of Switzerland was found to exhibit the best performance. With regard to the ratio of the number of publications to the Gross Domestic Product (GDP) per Capita, the USA remains the leading nation but countries like India and China with a low GDP and high population numbers are also profiting. Though the topic is a global matter, the scientific output on patient safety is centred mainly in industrialised countries. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

Research on the method of information system risk state estimation based on clustering particle filter

NASA Astrophysics Data System (ADS)

Cui, Jia; Hong, Bei; Jiang, Xuepeng; Chen, Qinghua

2017-05-01

With the purpose of reinforcing correlation analysis of risk assessment threat factors, a dynamic assessment method of safety risks based on particle filtering is proposed, which takes threat analysis as the core. Based on the risk assessment standards, the method selects threat indicates, applies a particle filtering algorithm to calculate influencing weight of threat indications, and confirms information system risk levels by combining with state estimation theory. In order to improve the calculating efficiency of the particle filtering algorithm, the k-means cluster algorithm is introduced to the particle filtering algorithm. By clustering all particles, the author regards centroid as the representative to operate, so as to reduce calculated amount. The empirical experience indicates that the method can embody the relation of mutual dependence and influence in risk elements reasonably. Under the circumstance of limited information, it provides the scientific basis on fabricating a risk management control strategy.

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

A cross-species translational pharmacokinetic-pharmacodynamic evaluation of core body temperature reduction by the TRPM8 blocker PF-05105679.

PubMed

Gosset, James R; Beaumont, Kevin; Matsuura, Tomomi; Winchester, Wendy; Attkins, Neil; Glatt, Sophie; Lightbown, Ian; Ulrich, Kristina; Roberts, Sonia; Harris, Jolie; Mesic, Emir; van Steeg, Tamara; Hijdra, Diana; van der Graaf, Piet H

2017-11-15

PF-05105679 is a moderately potent TRPM8 blocker which has been evaluated for the treatment of cold pain sensitivity. The TRPM8 channel is responsible for the sensation of cold environmental temperatures and has been implicated in regulation of core body temperature. Consequently, blockade of TRPM8 has been suggested to result in lowering of core body temperature. As part of the progression to human studies, the effect of PF-05105679 on core body temperature has been investigated in animals. Safety pharmacology studies showed that PF-05105679 reduced core body temperature in a manner that was inversely related to body weight of the species tested (greater exposure to PF-05105679 was required to lower temperature by 1°C in higher species). Based on an allometric (body weight) relationship, it was hypothesized that PF-05105679 would not lower core body temperature in humans at exposures that could exhibit pharmacological effects on cold pain sensation. On administration to humans, PF-05105679 was indeed effective at reversing the cold pain sensation associated with the cold pressor test in the absence of effects on core body temperature. Copyright © 2017 Elsevier B.V. All rights reserved.

Structural Element Testing in Support of the Design of the NASA Composite Crew Module

NASA Technical Reports Server (NTRS)

Kellas, Sotiris; Jackson, Wade C.; Thesken, John C.; Schleicher, Eric; Wagner, Perry; Kirsch, Michael T.

2012-01-01

In January 2007, the NASA Administrator and Associate Administrator for the Exploration Systems Mission Directorate chartered the NASA Engineering and Safety Center (NESC) to design, build, and test a full-scale Composite Crew Module (CCM). For the design and manufacturing of the CCM, the team adopted the building block approach where design and manufacturing risks were mitigated through manufacturing trials and structural testing at various levels of complexity. Following NASA's Structural Design Verification Requirements, a further objective was the verification of design analysis methods and the provision of design data for critical structural features. Test articles increasing in complexity from basic material characterization coupons through structural feature elements and large structural components, to full-scale structures were evaluated. This paper discusses only four elements tests three of which include joints and one that includes a tapering honeycomb core detail. For each test series included are specimen details, instrumentation, test results, a brief analysis description, test analysis correlation and conclusions.

Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

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

Low, J.O.; Schmitt, B.E.

1988-02-01

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may bemore » exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.« less

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

Rational design of novel, fluorescent, tagged glutamic acid dendrimers with different terminal groups and in silico analysis of their properties

PubMed Central

Martinho, Nuno; Silva, Liana C; Florindo, Helena F; Brocchini, Steve; Zloh, Mire; Barata, Teresa S

2017-01-01

Dendrimers are hyperbranched polymers with a multifunctional architecture that can be tailored for the use in various biomedical applications. Peptide dendrimers are particularly relevant for drug delivery applications due to their versatility and safety profile. The overall lack of knowledge of their three-dimensional structure, conformational behavior and structure–activity relationship has slowed down their development. Fluorophores are often conjugated to dendrimers to study their interaction with biomolecules and provide information about their mechanism of action at the molecular level. However, these probes can change dendrimer surface properties and have a direct impact on their interactions with biomolecules and with lipid membranes. In this study, we have used computer-aided molecular design and molecular dynamics simulations to identify optimal topology of a poly(l-glutamic acid) (PG) backbone dendrimer that allows incorporation of fluorophores in the core with minimal availability for undesired interactions. Extensive all-atom molecular dynamic simulations with the CHARMM force field were carried out for different generations of PG dendrimers with the core modified with a fluorophore (nitrobenzoxadiazole and Oregon Green 488) and various surface groups (glutamic acid, lysine and tryptophan). Analysis of structural and topological features of all designed dendrimers provided information about their size, shape, internal distribution and dynamic behavior. We have found that four generations of a PG dendrimer are needed to ensure minimal exposure of a core-conjugated fluorophore to external environment and absence of undesired interactions regardless of the surface terminal groups. Our findings suggest that NBD-PG-G4 can provide a suitable scaffold to be used for biophysical studies of surface-modified dendrimers to provide a deeper understanding of their intermolecular interactions, mechanisms of action and trafficking in a biological system. PMID:29026301

Rational design of novel, fluorescent, tagged glutamic acid dendrimers with different terminal groups and in silico analysis of their properties.

PubMed

Martinho, Nuno; Silva, Liana C; Florindo, Helena F; Brocchini, Steve; Zloh, Mire; Barata, Teresa S

2017-01-01

Dendrimers are hyperbranched polymers with a multifunctional architecture that can be tailored for the use in various biomedical applications. Peptide dendrimers are particularly relevant for drug delivery applications due to their versatility and safety profile. The overall lack of knowledge of their three-dimensional structure, conformational behavior and structure-activity relationship has slowed down their development. Fluorophores are often conjugated to dendrimers to study their interaction with biomolecules and provide information about their mechanism of action at the molecular level. However, these probes can change dendrimer surface properties and have a direct impact on their interactions with biomolecules and with lipid membranes. In this study, we have used computer-aided molecular design and molecular dynamics simulations to identify optimal topology of a poly(l-glutamic acid) (PG) backbone dendrimer that allows incorporation of fluorophores in the core with minimal availability for undesired interactions. Extensive all-atom molecular dynamic simulations with the CHARMM force field were carried out for different generations of PG dendrimers with the core modified with a fluorophore (nitrobenzoxadiazole and Oregon Green 488) and various surface groups (glutamic acid, lysine and tryptophan). Analysis of structural and topological features of all designed dendrimers provided information about their size, shape, internal distribution and dynamic behavior. We have found that four generations of a PG dendrimer are needed to ensure minimal exposure of a core-conjugated fluorophore to external environment and absence of undesired interactions regardless of the surface terminal groups. Our findings suggest that NBD-PG-G4 can provide a suitable scaffold to be used for biophysical studies of surface-modified dendrimers to provide a deeper understanding of their intermolecular interactions, mechanisms of action and trafficking in a biological system.

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

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

Hongbin Zhang; Haihua Zhao; Cliff Davis

2008-06-01

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

Strategies for enhancing perioperative safety: promoting joy and meaning in the workforce.

PubMed

Morath, Julianne; Filipp, Rhonda; Cull, Michael

2014-10-01

Workforce safety is a precondition of patient safety, and safety from both physical and psychological harm in the workplace is the foundation for an environment in which joy and meaning can exist. Achieving joy and meaning in the workplace allows health care workers to continuously improve the care they provide. This requires an environment in which disrespectful and harmful behaviors are not tolerated or ignored. Health care leaders have an obligation to create workplace cultures that are characterized by respect, transparency, accountability, learning, and quality care. Evidence suggests, however, that health care settings are rife with disrespectful behavior, poor teamwork, and unsafe working conditions. Solutions for addressing workplace safety problems include defining core values, tasking leaders to act as role models, and committing to becoming a high-reliability organization. Copyright © 2014 AORN, Inc. Published by Elsevier Inc. All rights reserved.

Indigenous Healing Knowledge and Infertility in Indonesia: Learning about Cultural Safety from Sasak Midwives.

PubMed

Bennett, Linda Rae

2017-01-01

In this article I demonstrate what can be learned from the indigenous healing knowledge and practices of traditional Sasak midwives on Lombok island in eastern Indonesia. I focus on the treatment of infertility, contrasting the differential experiences of Sasak women when they consult traditional midwives and biomedical doctors. Women's and midwives' perspectives provide critical insight into how cultural safety is both constituted and compromised in the context of reproductive health care. Core components of cultural safety embedded in the practices of traditional midwives include the treatment of women as embodied subjects rather than objectified bodies, and privileging physical contact as a healing modality. Cultural safety also encompasses respect for women's privacy and bodily dignity, as well as two-way and narrative communication styles. Local understandings of cultural safety have great potential to improve the routine practices of doctors, particularly in relation to doctor-patient communication and protocols for conducting pelvic exams.

How Pleasant Sounds Promote and Annoying Sounds Impede Health: A Cognitive Approach

PubMed Central

Andringa, Tjeerd C.; Lanser, J. Jolie L.

2013-01-01

This theoretical paper addresses the cognitive functions via which quiet and in general pleasurable sounds promote and annoying sounds impede health. The article comprises a literature analysis and an interpretation of how the bidirectional influence of appraising the environment and the feelings of the perceiver can be understood in terms of core affect and motivation. This conceptual basis allows the formulation of a detailed cognitive model describing how sonic content, related to indicators of safety and danger, either allows full freedom over mind-states or forces the activation of a vigilance function with associated arousal. The model leads to a number of detailed predictions that can be used to provide existing soundscape approaches with a solid cognitive science foundation that may lead to novel approaches to soundscape design. These will take into account that louder sounds typically contribute to distal situational awareness while subtle environmental sounds provide proximal situational awareness. The role of safety indicators, mediated by proximal situational awareness and subtle sounds, should become more important in future soundscape research. PMID:23567255

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

DOE PAGES

Lai, Shigang; Shi, Li; Fok, Alex; ...

2017-01-01

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

[METHOD OF INCREASING MICROBIOLOGICAL PURITY OF POWDER FROM COCOA-VELLA].

PubMed

Magomedov, G O; Cheremushkina, L V; Plotnikova, I V

2015-01-01

In the article there is described in detail the characteristic of the product of processing cocoa beans--cocoa-vella, there is presented a comparative analysis of the chemical composition, quality indices, the dispersive pattern, microbiological indices of the powder from the cocoa-vella in comparison to cocoa powder, obtained by traditional technology from the core of the cocoa beans. To improve the microbiological purity of the powder from the cocoa-vella there was suggested to be the modern and environmentally safe manner for the preparation of the powder The use of cocoa-vella disinfecting power by means of the electromagnetic field of ultrahigh frequency (RF EMF) was established to allow to obtain a product that meets the requirements of Technical Regulations of the Customs Union (TRCU 021/2011) on Food Safety. This work is of practical interest, since it helps to improve the safety of the powder from the cocoa-vella, and thus the quality of confectionery and food products based on it, which is relevant in terms of the management of a healthy diet.

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

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

Lai, Shigang; Shi, Li; Fok, Alex

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

Pore Water Transport of Enterococci out of Beach Sediments

PubMed Central

Phillips, Matthew C.; Solo-Gabriele, Helena M.; Reniers, Adrianus J. H. M.; Wang, John D.; Kiger, Russell T.; Abdel-Mottaleb, Noha

2011-01-01

Enterococci are used to evaluate the safety of beach waters and studies have identified beach sands as a source of these bacteria. In order to study and quantify the release of microbes from beach sediments, flow column systems were built to evaluate flow of pore water out of beach sediments. Results show a peak in enterococci (average of 10% of the total microbes in core) released from the sand core within one pore water volume followed by a marked decline to below detection. These results indicate that few enterococci are easily removed and that factors other than simple pore water flow control the release of the majority of enterococci within beach sediments. A significantly larger quantity and release of enterococci were observed in cores collected after a significant rain event suggesting the influx of fresh water can alter the release pattern as compared to cores with no antecedent rainfall. PMID:21945015

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Multi-Core Processor Memory Contention Benchmark Analysis Case Study

NASA Technical Reports Server (NTRS)

Simon, Tyler; McGalliard, James

2009-01-01

Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

Are HPV vaccination services accessible to high-risk communities? A spatial analysis of HPV-associated cancer and Chlamydia rates and safety-net clinics.

PubMed

Tsui, Jennifer; Rodriguez, Hector P; Gee, Gilbert C; Escobedo, Loraine A; Kominski, Gerald F; Bastani, Roshan

2013-12-01

While HPV vaccines can greatly benefit adolescents and young women from high-risk areas, little is known about whether safety-net immunization services are geographically accessible to communities at greatest risk for HPV-associated diseases. We explore the spatial relationship between areas with high HPV risk and proximity to safety-net clinics from an ecologic perspective. We used cancer registry data and Chlamydia surveillance data to identify neighborhoods within Los Angeles County with high risk for HPV-associated cancers. We examined proximity to safety-net clinics among neighborhoods with the highest risk. Proximity was measured as the shortest distance between each neighborhood center and the nearest clinic and having a clinic within 3 miles of each neighborhood center. The average 5-year non-age-adjusted rates were 1,940 cases per 100,000 for Chlamydia and 60 per 100,000 for HPV-associated cancers. A large majority, 349 of 386 neighborhoods with high HPV-associated cancer rates and 532 of 537 neighborhoods with high Chlamydia rates, had a clinic within 3 miles of the neighborhood center. Clinics were more likely to be located within close proximity to high-risk neighborhoods in the inner city. High-risk neighborhoods outside of this urban core area were less likely to be near accessible clinics. The majority of high-risk neighborhoods were geographically near safety-net clinics with HPV vaccination services. Due to low rates of vaccination, these findings suggest that while services are geographically accessible, additional efforts are needed to improve uptake. Programs aimed to increase awareness about the vaccine and to link underserved groups to vaccination services are warranted.

Identifying nontechnical skills associated with safety in the emergency department: a scoping review of the literature.

PubMed

Flowerdew, Lynsey; Brown, Ruth; Vincent, Charles; Woloshynowych, Maria

2012-05-01

Understanding the nontechnical skills specifically applicable to the emergency department (ED) is essential to facilitate training and more broadly consider interventions to reduce error. The aim of this scoping review is to first identify and then explore in depth the nontechnical skills linked to safety in the ED. The review was conducted in 2 stages. In stage 1, online databases were searched for published empirical studies linking nontechnical skills to safety and performance in the ED. Articles were analyzed to identify key ED nontechnical skills. In stage 2, these key skills were used to generate additional key words, which enabled a second search of the literature to be undertaken and expand on the evidence available for review. In stage 1, 11 articles were retrieved for data analysis and 9 core emergency medicine nontechnical skills were identified. These were communicating, managing workload, anticipating, situational awareness, supervising and providing feedback, leadership, maintaining standards, using assertiveness, and decisionmaking. In stage 2, a secondary search, using these 9 skills and related terms, uncovered a further 21 relevant articles. Therefore, 32 articles were used to describe the main nontechnical skills linked to safety in the ED. This article highlights the challenges of reviewing a topic for which the terms are not clearly defined in the literature. A novel methodological approach is described that provides a structured and transparent process for reviewing the literature in emerging areas of interest. A series of literature reviews focusing on individual nontechnical skills will provide a clearer understanding of how the skills identified contribute to safety in the ED. Copyright © 2011 American College of Emergency Physicians. Published by Mosby, Inc. All rights reserved.

Perceptions of health care professionals on the safety and security at Odi District Hospital, Gauteng, South Africa

PubMed Central

Okeke, Sunday O.

2017-01-01

Background For optimum delivery of service, an establishment needs to ensure a safe and secure environment. In 2011, the South African government promulgated the National Core Standards for Health Establishments for safety and security for all employees in all establishments. Little is known about whether these standards are being complied to. Aim and setting: To assess the perceptions of health care professionals (HCPs) on safety and security at Odi District Hospital. Methodology A sample of 181 out of a total of 341 HCPs was drawn through a systematic sampling method from each HCP category. Data were collected through a self-administered questionnaire. The SPSS® statistical software version 22 was used for data analysis. The level of statistical significance was set at < 0.05. Results There were more female respondents than male respondents (136; 75.10%). The dominant age group was 28–47 years (114; 57.46%). Perceptions on security personnel, their efficiency and the security system were significantly affirmed (p = 0.0001). The hospital infrastructure, surroundings and plan in emergencies were perceived to be safe (p < 0.0001). The hospital lighting system was perceived as inadequate (p = 0.0041). Only 36 (20.2%) HCPs perceived that hospital authorities were concerned about employees’ safety (p < 0.0001). Conclusion HCPs had positive perceptions regarding the hospital’s security system. Except for the negative perceptions of the lighting system and the perceived lack of hospital authorities’ concern for staff safety, perceptions of the HCPs on the hospital working environment were positive. The hospital authorities need to establish the basis of negative perceptions and enforce remedial measures to redress them. PMID:29113444

Are HPV vaccination services accessible to high-risk communities?: A spatial analysis of HPV-associated cancer and Chlamydia rates and safety-net clinics

PubMed Central

Tsui, Jennifer; Rodriguez, Hector P.; Gee, Gilbert C.; Escobedo, Loraine A.; Kominski, Gerald F.; Bastani, Roshan

2013-01-01

Purpose While HPV vaccines can greatly benefit adolescents and young women from high-risk areas, little is known about whether safety-net immunization services are geographically accessible to communities at greatest risk for HPV-associated diseases. We explore the spatial relationship between areas with high HPV risk and proximity to safety-net clinics from an ecologic perspective. Methods We used cancer registry data and Chlamydia surveillance data to identify neighborhoods within Los Angeles County with high risk for HPV-associated cancers. We examined proximity to safety-net clinics among neighborhoods with the highest risk. Proximity was measured as the shortest distance between each neighborhood center and the nearest clinic and having a clinic within 3 miles of each neighborhood center. Results The average 5-year non-age-adjusted rates were 1,940 cases per 100,000 for Chlamydia and 60 per 100,000 for HPV-associated cancers. A large majority, 349 of 386 neighborhoods with high HPV-associated cancer rates and 532 of 537 neighborhoods with high Chlamydia rates had a clinic within 3 miles of the neighborhood center. Clinics were more likely to be located within close proximity to high-risk neighborhoods in the inner city. High-risk neighborhoods outside of this urban core area were less likely to be near accessible clinics. Conclusions The majority of high-risk neighborhoods were geographically near safety-net clinics with HPV vaccination services. Due to low rates of vaccination, these findings suggest that while services are geographically accessible, additional efforts are needed to improve uptake. Programs aimed to increase awareness about the vaccine and to link underserved groups to vaccination services are warranted. PMID:24043448

Application of reliability-centered-maintenance to BWR ECCS motor operator valve performance

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

Feltus, M.A.; Choi, Y.A.

1993-01-01

This paper describes the application of reliability-centered maintenance (RCM) methods to plant probabilistic risk assessment (PRA) and safety analyses for four boiling water reactor emergency core cooling systems (ECCSs): (1) high-pressure coolant injection (HPCI); (2) reactor core isolation cooling (RCIC); (3) residual heat removal (RHR); and (4) core spray systems. Reliability-centered maintenance is a system function-based technique for improving a preventive maintenance program that is applied on a component basis. Those components that truly affect plant function are identified, and maintenance tasks are focused on preventing their failures. The RCM evaluation establishes the relevant criteria that preserve system function somore » that an RCM-focused approach can be flexible and dynamic.« less

Enhanced performance of core-shell structured polyaniline at helical carbon nanotube hybrids for ammonia gas sensor

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

Tian, Xin; Wang, Qiang; Chen, Xiangnan

2014-11-17

A core-shell structured hybrid of polyaniline at helical carbon nanotubes was synthesized using in situ polymerization, which the helical carbon nanotubes were uniformly surrounded by a layer of polyaniline nanorods array. More interestingly, repeatable responses were experimentally observed that the sensitivity to ammonia gas of the as-prepared helical shaped core-shell hybrid displays an enhancement of more than two times compared to those of only polyaniline or helical carbon nanotubes sensors because of the peculiar structures with high surface area. This kind of hybrid comprising nanorod arrays of conductive polymers covering carbon nanotubes and related structures provide a potential in sensorsmore » of trace gas detection for environmental monitoring and safety forecasting.« less

Fire Management/Suppression Systems/Concepts Relating to Aircraft Cabin Fire Safety.

DTIC Science & Technology

1982-07-01

polyamide- cm; cell size, cm; paper honeycomb; paper honeycomb; density, kg/cm 3 ; filler 2.413;0.31; 48.06; 2.413;0.31; 48.06; no core filler no core...0 4’ 14 ~ 1 4 44 N4 El4 (36M -11 - 0 0 N4 8A 0o N. 0 04 0n 0m C,0 0 44 . . 000040 -t 0 0 880 -8 ,0 04 0 C, 0 C4 4-0 OD -CD0C, m. g G - C4 A - 40 a

Effect of STOP technique on safety climate in a construction company.

PubMed

Darvishi, Ebrahim; Maleki, Afshin; Dehestaniathar, Saeed; Ebrahemzadih, Mehrzad

2015-01-01

Safety programs are a core part of safety management in workplaces that can reduce incidents and injuries. The aim of this study was to investigate the influence of Safety Training Observation Program (STOP) technique as a behavior modification program on safety climate in a construction company. This cross-sectional study was carried out on workers of the Petrochemical Construction Company, western Iran. In order to improve safety climate, an unsafe behavior modification program entitled STOP was launched among workers of project during 12 months from April 2013 and April 2014. The STOP technique effectiveness in creating a positive safety climate was evaluated using the Safety Climate Assessment Toolkit. 76.78% of total behaviors were unsafe. 54.76% of total unsafe acts/ at-risk behaviors were related to the fall hazard. The most cause of unsafe behaviors was associated with habit and unavailability of safety equipment. After 12 month of continuous implementation the STOP technique, 55.8% of unsafe behaviors reduced among workers. The average score of safety climate evaluated using of the Toolkit, before and after the implementation of the STOP technique was 5.77 and 7.24, respectively. The STOP technique can be considered as effective approach for eliminating at-risk behavior, reinforcing safe work practices, and creating a positive safety climate in order to reduction incidents/injuries.

Balancing intertwined responsibilities: A grounded theory study of teamwork in everyday intensive care unit practice.

PubMed

Bjurling-Sjöberg, Petronella; Wadensten, Barbro; Pöder, Ulrika; Jansson, Inger; Nordgren, Lena

2017-03-01

This study aimed to describe and explain teamwork and factors that influence team processes in everyday practice in an intensive care unit (ICU) from a staff perspective. The setting was a Swedish ICU. Data were collected from 38 ICU staff in focus groups with registered nurses, assistant nurses, and anaesthetists, and in one individual interview with a physiotherapist. Constant comparative analysis according to grounded theory was conducted, and to identify the relations between the emerged categories, the paradigm model was applied. The core category to emerge from the data was "balancing intertwined responsibilities." In addition, eleven categories that related to the core category emerged. These categories described and explained the phenomenon's contextual conditions, causal conditions, and intervening conditions, as well as the staff actions/interactions and the consequences that arose. The findings indicated that the type of teamwork fluctuated due to circumstantial factors. Based on the findings and on current literature, strategies that can optimise interprofessional teamwork are presented. The analysis generated a conceptual model, which aims to contribute to existing frameworks by adding new dimensions about perceptions of team processes within an ICU related to staff actions/interactions. This model may be utilised to enhance the understanding of existing contexts and processes when designing and implementing interventions to facilitate teamwork in the pursuit of improving healthcare quality and patient safety.

A new safety channel based on ¹⁷N detection in research reactors.

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

Amended Safety Assessment of Isethionate Salts as Used in Cosmetics.

PubMed

Burnett, Christina L; Heldreth, Bart; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan

The Cosmetic Ingredient Review (CIR) Expert Panel (Panel) rereviewed the safety of 12 isethionate salts as used in cosmetics and concluded that these ingredients are safe in the present practices of use and concentration, when formulated to be nonirritating. These isethionate salts are reported to function mostly as surfactants and cleansing agents in cosmetic products. The Panel reviewed the available animal and clinical data as well as information from previous CIR reports. Although there are data gaps, the shared chemical core structure, expected similarities in physicochemical properties, and similar functions and concentrations in cosmetics enabled grouping these ingredients and reading across the available toxicological data to support the safety assessment of each ingredient.

Sientra portfolio of Silimed brand shaped implants with high-strength silicone gel: a 5-year primary augmentation clinical study experience and a postapproval experience-results from a single-surgeon 108-patient series.

PubMed

Haws, Melinda J; Schwartz, Michael R; Berger, Lewis H; Daulton, Kimber L

2014-07-01

The Sientra portfolio of silicone gel breast implants was approved by the Food and Drug Administration on March 9, 2012, and included the first approved shaped implants in the United States. The 5-year results from Sientra's Core Gel and Continued Access Study and the results of a single surgeon are presented. This analysis used the data of 640 shaped implants in 321 primary augmentation patients implanted by 16 study surgeons through 5 years. The Kaplan-Meier method was used to analyze safety endpoints. In addition, analysis is presented for a single surgeon's results of 213 shaped implants in 108 postapproval patients through up to 16 months of follow-up (9-month mean) using a separate frequency analysis. The overall risk of rupture for primary augmentation patients through 5 years was 0.4%, the risk of infection was 1.4%, and the risk of capsular contracture (Baker grade III/IV) was 3.9%. Reported surgeon satisfaction was 100%, and patient satisfaction remained high. In the separate single-surgeon analysis, after 16 months, 4 of the 108 patients experienced a complication (3.7%) and 3 underwent a reoperation (2.8%). Complications included infection, ptosis (0.9%, each), and capsular contracture (1.9%). The results of Sientra's large clinical study and the postapproval data from a single surgeon demonstrate the safety and effectiveness of Sientra's shaped implants. The review of the data and author's experience illustrate the ease of incorporating shaped implants into any surgical practice.

A streamlined failure mode and effects analysis.

PubMed

Ford, Eric C; Smith, Koren; Terezakis, Stephanie; Croog, Victoria; Gollamudi, Smitha; Gage, Irene; Keck, Jordie; DeWeese, Theodore; Sibley, Greg

2014-06-01

Explore the feasibility and impact of a streamlined failure mode and effects analysis (FMEA) using a structured process that is designed to minimize staff effort. FMEA for the external beam process was conducted at an affiliate radiation oncology center that treats approximately 60 patients per day. A structured FMEA process was developed which included clearly defined roles and goals for each phase. A core group of seven people was identified and a facilitator was chosen to lead the effort. Failure modes were identified and scored according to the FMEA formalism. A risk priority number,RPN, was calculated and used to rank failure modes. Failure modes with RPN > 150 received safety improvement interventions. Staff effort was carefully tracked throughout the project. Fifty-two failure modes were identified, 22 collected during meetings, and 30 from take-home worksheets. The four top-ranked failure modes were: delay in film check, missing pacemaker protocol/consent, critical structures not contoured, and pregnant patient simulated without the team's knowledge of the pregnancy. These four failure modes had RPN > 150 and received safety interventions. The FMEA was completed in one month in four 1-h meetings. A total of 55 staff hours were required and, additionally, 20 h by the facilitator. Streamlined FMEA provides a means of accomplishing a relatively large-scale analysis with modest effort. One potential value of FMEA is that it potentially provides a means of measuring the impact of quality improvement efforts through a reduction in risk scores. Future study of this possibility is needed.

Information collection and processing of dam distortion in digital reservoir system

NASA Astrophysics Data System (ADS)

Liang, Yong; Zhang, Chengming; Li, Yanling; Wu, Qiulan; Ge, Pingju

2007-06-01

The "digital reservoir" is usually understood as describing the whole reservoir with digital information technology to make it serve the human existence and development furthest. Strictly speaking, the "digital reservoir" is referred to describing vast information of the reservoir in different dimension and space-time by RS, GPS, GIS, telemetry, remote-control and virtual reality technology based on computer, multi-media, large-scale memory and wide-band networks technology for the human existence, development and daily work, life and entertainment. The core of "digital reservoir" is to realize the intelligence and visibility of vast information of the reservoir through computers and networks. The dam is main building of reservoir, whose safety concerns reservoir and people's safety. Safety monitoring is important way guaranteeing the dam's safety, which controls the dam's running through collecting the dam's information concerned and developing trend. Safety monitoring of the dam is the process from collection and processing of initial safety information to forming safety concept in the brain. The paper mainly researches information collection and processing of the dam by digital means.

Pasireotide can induce sustained decreases in urinary cortisol and provide clinical benefit in patients with Cushing's disease: results from an open-ended, open-label extension trial.

PubMed

Schopohl, Jochen; Gu, Feng; Rubens, Robert; Van Gaal, Luc; Bertherat, Jérôme; Ligueros-Saylan, Monica; Trovato, Andrew; Hughes, Gareth; Salgado, Luiz R; Boscaro, Marco; Pivonello, Rosario

2015-10-01

Report the efficacy and safety of pasireotide sc in patients with Cushing's disease during an open-ended, open-label extension to a randomized, double-blind, 12-month, Phase III study. 162 patients entered the core study. 58 patients who had mean UFC ≤ ULN at month 12 or were benefiting clinically from pasireotide entered the extension. Patients received the same dose of pasireotide as at the end of the core study (300-1,200 μg bid). Dose titration was permitted according to efficacy or drug-related adverse events. 40 patients completed 24 months' treatment. Of the patients who entered the extension, 50.0% (29/58) and 34.5% (20/58) had controlled UFC (UFC ≤ ULN) at months 12 and 24, respectively. The mean percentage decrease in UFC was 57.3% (95% CI 40.7-73.9; n = 52) and 62.1% (50.8-73.5; n = 33) after 12 and 24 months' treatment, respectively. Improvements in clinical signs of Cushing's disease were sustained up to month 24. The most frequent drug-related adverse events in patients who received ≥1 dose of pasireotide (n = 162) from core baseline until the 24-month cut-off were diarrhea (55.6%), nausea (48.1%), hyperglycemia (38.9%), and cholelithiasis (31.5%). No new safety issues were identified during the extension. Reductions in mean UFC and improvements in clinical signs of Cushing's disease were maintained over 24 months of pasireotide treatment. The safety profile of pasireotide is typical for a somatostatin analogue, except for the frequency and degree of hyperglycemia; patients should be monitored for changes in glucose homeostasis. Pasireotide represents the first approved pituitary-targeted treatment for patients with Cushing's disease.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

10 CFR 830.204 - Documented safety analysis.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Documented safety analysis. 830.204 Section 830.204 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.204 Documented safety analysis... approval from DOE for the methodology used to prepare the documented safety analysis for the facility...

Analysis of the TREAT LEU Conceptual Design

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Managementmore » and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO 2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.« less

Interpretation of the results of the CORA-33 dry core BWR test

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

Ott, L.J.; Hagen, S.

All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions ofmore » a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.« less

Potential Application of a Thermoelectric Generator in Passive Cooling System of Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Wang, Dongqing; Liu, Yu; Jiang, Jin; Pang, Wei; Lau, Woon Ming; Mei, Jun

2017-05-01

In the design of nuclear power plants, various natural circulation passive cooling systems are considered to remove residual heat from the reactor core in the event of a power loss and maintain the plant's safety. These passive systems rely on gravity differences of fluids, resulting from density differentials, rather than using an external power-driven system. Unfortunately, a major drawback of such systems is their weak driving force, which can negatively impact safety. In such systems, there is a temperature difference between the heat source and the heat sink, which potentially offers a natural platform for thermoelectric generator (TEG) applications. While a previous study designed and analyzed a TEG-based passive core cooling system, this paper considers TEG applications in other passive cooling systems of nuclear power plants, after which the concept of a TEG-based passive cooling system is proposed. In such a system, electricity is produced using the system's temperature differences through the TEG, and this electricity is used to further enhance the cooling process.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

1986-01-01

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

1986-07-01

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

[Process orientation as a tool of strategic approaches to corporate governance and integrated management systems].

PubMed

Sens, Brigitte

2010-01-01

The concept of general process orientation as an instrument of organisation development is the core principle of quality management philosophy, i.e. the learning organisation. Accordingly, prestigious quality awards and certification systems focus on process configuration and continual improvement. In German health care organisations, particularly in hospitals, this general process orientation has not been widely implemented yet - despite enormous change dynamics and the requirements of both quality and economic efficiency of health care processes. But based on a consistent process architecture that considers key processes as well as management and support processes, the strategy of excellent health service provision including quality, safety and transparency can be realised in daily operative work. The core elements of quality (e.g., evidence-based medicine), patient safety and risk management, environmental management, health and safety at work can be embedded in daily health care processes as an integrated management system (the "all in one system" principle). Sustainable advantages and benefits for patients, staff, and the organisation will result: stable, high-quality, efficient, and indicator-based health care processes. Hospitals with their broad variety of complex health care procedures should now exploit the full potential of total process orientation. Copyright © 2010. Published by Elsevier GmbH.

Effects of heated hydrotherapy on muscle HSP70 and glucose metabolism in old and young vervet monkeys.

PubMed

Kavanagh, Kylie; Davis, Ashely T; Jenkins, Kurt A; Flynn, D Mickey

2016-07-01

Increasing heat shock protein 70 (HSP70) in aged and/or insulin-resistant animal models confers benefits to healthspan and lifespan. Heat application to increase core temperature induces HSPs in metabolically important tissues, and preliminary human and animal data suggest that heated hydrotherapy is an effective method to achieve increased HSPs. However, safety concerns exist, particularly in geriatric medicine where organ and cardiovascular disease commonly will preexist. We evaluated young vervet monkeys compared to old, insulin-resistant vervet monkeys (Chlorocebus aethiops sabaeus) in their core temperatures, glucose tolerance, muscle HSP70 level, and selected safety biomarkers after 10 sessions of hot water immersions administered twice weekly. Hot water immersion robustly induced the heat shock response in muscles. We observed that heat-treated old and young monkeys have significantly higher muscle HSP70 than control monkeys and treatment was without significant adverse effects on organ or cardiovascular health. Heat therapy improved pancreatic responses to glucose challenge and tended to normalize glucose excursions. A trend for worsened blood pressure and glucose values in the control monkeys and improved values in heat-treated monkeys were seen to support further investigation into the safety and efficacy of this intervention for metabolic syndrome or diabetes in young or old persons unable to exercise.

Budesonide Multi-matrix for the Treatment of Patients with Ulcerative Colitis.

PubMed

Lichtenstein, Gary R

2016-02-01

Ulcerative colitis (UC) is a chronic idiopathic inflammatory disorder in which patients cycle between active disease and remission. Budesonide multi-matrix (MMX) is an oral second-generation corticosteroid designed to deliver active drug throughout the colon. In pharmacokinetic studies, the mean relative absorption of budesonide in the region between the ascending colon and the descending/sigmoid colon was 95.9 %. In 2 identically designed, phase 3 studies (CORE I and II), budesonide MMX 9 mg once daily was efficacious and well tolerated for induction of remission of mild to moderate UC. Clinical and endoscopic remission rates were 17.9 % (CORE I) and 17.4 % (CORE II) for budesonide MMX 9 mg compared with 7.4 and 4.5 %, respectively, with placebo (p < 0.05, budesonide MMX 9 mg vs. placebo in both studies), 12.1 % with mesalamine 2.4 g, and 12.6 % with budesonide controlled ileal release capsules 9 mg. A 12-month maintenance therapy study suggested that budesonide MMX 6 mg may prolong time to clinical relapse: Median time was >1 year with budesonide MMX 6 mg versus 181 days (p = 0.02) with placebo; however, further studies are needed. In the CORE studies, budesonide MMX exhibited a favorable safety profile; the majority of adverse events were mild or moderate in intensity, and serious adverse events were uncommon. Furthermore, rates of potential glucocorticoid-related adverse events were comparable across treatment groups. The long-term (12-month) safety of budesonide MMX appears to be comparable with placebo. Data support budesonide MMX in the management algorithm of UC.

New perspectives on occupational health and safety in immigrant populations: studying the intersection between immigrant background and gender.

PubMed

Mousaid, Sarah; De Moortel, Deborah; Malmusi, Davide; Vanroelen, Christophe

2016-01-01

Few studies investigating health inequalities pay attention to the intersection between several social determinants of health. The purpose of this article is to examine the relation between perceptions of work-related health and safety risk (WHSR) and (1) immigrant background and (2) gender in the EU-15. The effects are controlled for educational attainment, the quality of work (QOW) and occupation. Pooled data from the European Social Survey 2004 and 2010 are used in this study. The sample is restricted to respondents of working age (16-65 years) (N = 17,468). The immigrants are divided into two groups according to their country of origin: (semi-)periphery and core countries. Both groups of immigrants are compared to natives. Additionally, the research population is stratified by gender. Descriptive statistics and logistic regression analyses are used. Core immigrants (both men and women) do not differ from natives in terms of QOW. (Semi-)periphery immigrants (both men and women) are employed in jobs with lower QOW. While no differences in WHSR are found among men, female immigrants (both (semi-)periphery and core) have significantly more WHSR compared to native women. Although WHSR is generally lower in women, (semi-)periphery women have a similar prevalence of WHSR as men. (Semi-)periphery immigrants are employed in lower quality jobs, while core immigrants do not differ from natives in that regard. Female immigrant workers--especially those from (semi-)periphery countries--have higher WHSR compared to native women. Our findings highlight the importance of an intersectional approach in the study of work-related health inequalities.

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Promoting a Culture of Safety as a Patient Safety Strategy

PubMed Central

Weaver, Sallie J.; Lubomksi, Lisa H.; Wilson, Renee F.; Pfoh, Elizabeth R.; Martinez, Kathryn A.; Dy, Sydney M.

2015-01-01

Developing a culture of safety is a core element of many efforts to improve patient safety and care quality. This systematic review identifies and assesses interventions used to promote safety culture or climate in acute care settings. The authors searched MEDLINE, CINAHL, PsycINFO, Cochrane, and EMBASE to identify relevant English-language studies published from January 2000 to October 2012. They selected studies that targeted health care workers practicing in inpatient settings and included data about change in patient safety culture or climate after a targeted intervention. Two raters independently screened 3679 abstracts (which yielded 33 eligible studies in 35 articles), extracted study data, and rated study quality and strength of evidence. Eight studies included executive walk rounds or interdisciplinary rounds; 8 evaluated multicomponent, unit-based interventions; and 20 included team training or communication initiatives. Twenty-nine studies reported some improvement in safety culture or patient outcomes, but measured outcomes were highly heterogeneous. Strength of evidence was low, and most studies were pre–post evaluations of low to moderate quality. Within these limits, evidence suggests that interventions can improve perceptions of safety culture and potentially reduce patient harm. PMID:23460092

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

The social practice of rescue: the safety implications of acute illness trajectories and patient categorisation in medical and maternity settings.

PubMed

Mackintosh, Nicola; Sandall, Jane

2016-02-01

The normative position in acute hospital care when a patient is seriously ill is to resuscitate and rescue. However, a number of UK and international reports have highlighted problems with the lack of timely recognition, treatment and referral of patients whose condition is deteriorating while being cared for on hospital wards. This article explores the social practice of rescue, and the structural and cultural influences that guide the categorisation and ordering of acutely ill patients in different hospital settings. We draw on Strauss et al.'s notion of the patient trajectory and link this with the impact of categorisation practices, thus extending insights beyond those gained from emergency department triage to care management processes further downstream on the hospital ward. Using ethnographic data collected from medical wards and maternity care settings in two UK inner city hospitals, we explore how differences in population, cultural norms, categorisation work and trajectories of clinical deterioration interlink and influence patient safety. An analysis of the variation in findings between care settings and patient groups enables us to consider socio-political influences and the specifics of how staff manage trade-offs linked to the enactment of core values such as safety and equity in practice. © 2015 The Authors. Sociology of Health & Illness published by John Wiley & Sons Ltd on behalf of Foundation for SHIL.

A review for identification of initiating events in event tree development process on nuclear power plants

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

Riyadi, Eko H., E-mail: e.riyadi@bapeten.go.id

2014-09-30

Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logicmore » model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.« less

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

Grabaskas, David; Brunett, Acacia J.; Passerini, Stefano

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory (Argonne) participated in a two year collaboration to modernize and update the probabilistic risk assessment (PRA) for the PRISM sodium fast reactor. At a high level, the primary outcome of the project was the development of a next-generation PRA that is intended to enable risk-informed prioritization of safety- and reliability-focused research and development. A central Argonne task during this project was a reliability assessment of passive safety systems, which included the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks of the metal fuel core. Both systems were examinedmore » utilizing a methodology derived from the Reliability Method for Passive Safety Functions (RMPS), with an emphasis on developing success criteria based on mechanistic system modeling while also maintaining consistency with the Fuel Damage Categories (FDCs) of the mechanistic source term assessment. This paper provides an overview of the reliability analyses of both systems, including highlights of the FMEAs, the construction of best-estimate models, uncertain parameter screening and propagation, and the quantification of system failure probability. In particular, special focus is given to the methodologies to perform the analysis of uncertainty propagation and the determination of the likelihood of violating FDC limits. Additionally, important lessons learned are also reviewed, such as optimal sampling methodologies for the discovery of low likelihood failure events and strategies for the combined treatment of aleatory and epistemic uncertainties.« less

EVALUATION OF THE HTA CORE MODEL FOR NATIONAL HEALTH TECHNOLOGY ASSESSMENT REPORTS: COMPARATIVE STUDY AND EXPERIENCES FROM EUROPEAN COUNTRIES.

PubMed

Kõrge, Kristina; Berndt, Nadine; Hohmann, Juergen; Romano, Florence; Hiligsmann, Mickael

2017-01-01

The health technology assessment (HTA) Core Model® is a tool for defining and standardizing the elements of HTA analyses within several domains for producing structured reports. This study explored the parallels between the Core Model and a national HTA report. Experiences from various European HTA agencies were also investigated to determine the Core Model's adaptability to national reports. A comparison between a national report on Genetic Counseling, produced by the Cellule d'expertise médicale Luxembourg, and the Core Model was performed to identify parallels in terms of relevant and comparable assessment elements (AEs). Semi-structured interviews with five representatives from European HTA agencies were performed to assess their user experiences with the Core Model. The comparative study revealed that 50 percent of the total number (n = 144) of AEs in the Core Model were relevant for the national report. Of these 144 AEs from the Core Model, 34 (24 percent) were covered in the national report. Some AEs were covered only partly. The interviewees emphasized flexibility in using the Core Model and stated that the most important aspects to be evaluated include characteristics of the disease and technology, clinical effectiveness, economic aspects, and safety. In the present study, the national report covered an acceptable number of AEs of the Core Model. These results need to be interpreted with caution because only one comparison was performed. The Core Model can be used in a flexible manner, applying only those elements that are relevant from the perspective of the technology assessment and specific country context.

Effectiveness and efficiency of different weight machine-based strength training programmes for patients with hip or knee osteoarthritis: a protocol for a quasi-experimental controlled study in the context of health services research.

PubMed

Krauss, Inga; Müller, Gerhard; Steinhilber, Benjamin; Haupt, Georg; Janssen, Pia; Martus, Peter

2017-01-01

Osteoarthritis is a chronic musculoskeletal disease with a major impact on the individual and the healthcare system. As there is no cure, therapy aims for symptom release and reduction of disease progression. Physical exercises have been defined as a core treatment for osteoarthritis. However, research questions related to dose response, sustainability of effects, economic efficiency and safety are still open and will be evaluated in this trial, investigating a progressive weight machine-based strength training. This is a quasi-experimental controlled trial in the context of health services research. The intervention group (n=300) is recruited from participants of an offer for insurants of a health insurance company suffering from hip or knee osteoarthritis. Potential participants of the control group are selected and written to from the insurance database according to predefined matching criteria. The final statistical twins from the control responders will be determined via propensity score matching (n=300). The training intervention comprises 24 supervised mandatory sessions (2/week) and another 12 facultative sessions (1/week). Exercises include resistance training for the lower extremity and core muscles by use of weight machines and small training devices. The training offer is available at two sites. They differ with respect to the weight machines in use resulting in different dosage parameters. Primary outcomes are self-reported pain and function immediately after the 12-week intervention period. Health-related quality of life, self-efficacy, cost utility and safety will be evaluated as secondary outcomes. Secondary analysis will be undertaken with two strata related to study site. Participants will be followed up 6, 12 and 24 months after baseline. German Clinical Trial Register DRKS00009257. Pre-results.

Two-Port Pars Plana Anterior and Central Core Vitrectomy (Lam Floaterectomy) in Combination With Phacoemulsification and Intraocular Lens Implantation Under Topical Anesthesia for Patients with Cataract and Significant Floaters: Results of the First 50 Consecutive Cases.

PubMed

Lam, Dennis S C; Leung, Hiu Ying; Liu, Shu; Radke, Nishant; Yuan, Ye; Lee, Vincent Y W

2017-01-01

To study the safety and efficacy of 2-port pars plana anterior and central core vitrectomy (Lam floaterectomy) in combination with phacoemulsification (phaco) and intraocular lens implantation (IOL) for patients with cataract and significant floaters under topical anesthesia. Retrospective review of the first 50 consecutive cases. A standardized treatment protocol was used for patients with cataract and significant (moderate to severe) floaters (duration > 3 months). Data analysis included intraoperative and postoperative complications, floater status, and patient satisfaction. There were 50 eyes (38 patients) with a male-to-female ratio of 1 to 2.3. Twelve patients had bilateral eye surgeries. Mean age was 58.10 ± 9.85 years (range, 39-83). All patients completed the 3-month follow-up. One eye had mild vitreous hemorrhage at the end of surgery arising from sclerotomy wound oozing. No other intraoperative compli-cations were encountered. Postoperatively, there was 1 case of transient hypotony and 1 case of congestion at sclerotomy wound. No cases of retinal break or detachment, or clinically significant macular edema, were reported. There were 5 cases (10%) of mild residual floaters and 1 case (2%) of floater recurrence. Total floater clearance rate was 88%. Patient satisfaction rates were 80%, 14%, 6%, and 0% for very satisfied, satis-fied, acceptable, and unsatisfied, respectively. The 3-month results in terms of safety and efficacy of the Lam floaterectomy in combination with phaco and IOLfor patients with cataract and significant floaters under topical anesthesia are encouraging. Further larger-scale, prospective, multicenter studies seem warranted. Copyright© 2017 Asia-Pacific Academy of Ophthalmology.

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

Dionne, B.; Tzanos, C. P.

To support the safety analyses required for the conversion of the Belgian Reactor 2 (BR2) from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, the simulation of a number of loss-of-flow tests, with or without loss of pressure, has been undertaken. These tests were performed at BR2 in 1963 and used instrumented fuel assemblies (FAs) with thermocouples (TC) imbedded in the cladding as well as probes to measure the FAs power on the basis of their coolant temperature rise. The availability of experimental data for these tests offers an opportunity to better establish the credibility of the RELAP5-3D model andmore » methodology used in the conversion analysis. In order to support the HEU to LEU conversion safety analyses of the BR2 reactor, RELAP simulations of a number of loss-of-flow/loss-of-pressure tests have been undertaken. Preliminary analyses showed that the conservative power distributions used historically in the BR2 RELAP model resulted in a significant overestimation of the peak cladding temperature during the transient. Therefore, it was concluded that better estimates of the steady-state and decay power distributions were needed to accurately predict the cladding temperatures measured during the tests and establish the credibility of the RELAP model and methodology. The new approach ('best estimate' methodology) uses the MCNP5, ORIGEN-2 and BERYL codes to obtain steady-state and decay power distributions for the BR2 core during the tests A/400/1, C/600/3 and F/400/1. This methodology can be easily extended to simulate any BR2 core configuration. Comparisons with measured peak cladding temperatures showed a much better agreement when power distributions obtained with the new methodology are used.« less

Long-term follow-up of a phase 2 study of oral teriflunomide in relapsing multiple sclerosis: safety and efficacy results up to 8.5 years

PubMed Central

Li, David K; Freedman, Mark S; Truffinet, Philippe; Benzerdjeb, Hadj; Wang, Dazhe; Bar-Or, Amit; Traboulsee, Anthony L; Reiman, Lucy E; O’Connor, Paul W

2012-01-01

Background: Teriflunomide, an oral disease-modifying therapy in development for patients with relapsing forms of multiple sclerosis (RMS), was well tolerated and effective in reducing magnetic resonance imaging (MRI) lesions in 179 RMS patients in a phase 2 36-week, placebo-controlled study. Methods: A total of 147 patients who completed the core study entered an open-label extension. Teriflunomide patients continued their assigned dose, and placebo patients were re-allocated to teriflunomide, 7 mg/day or 14 mg/day. An interim analysis was performed at a cut-off on January 8 2010. Results: The mean and median duration of study treatment, including both the core and extension phase, from baseline to the interim cut-off, was 5.6 years (standard deviation: 2.7 years) and 7.1 years (range: 0.05–8.5 years), respectively. Of 147 patients, 62 (42.2%) discontinued (19% due to treatment-emergent adverse events (TEAEs)). The most common TEAEs were mild infections, fatigue, sensory disturbances and diarrhoea. No serious opportunistic infections occurred, with no discontinuations due to infection. Asymptomatic alanine aminotransferase increases (≤3× upper limit of normal (ULN)) were common (7 mg, 64.2%; 14 mg, 62.1%); increases >3×ULN were similar across groups (7 mg, 12.3%; 14 mg, 12.1%). Mild decreases in neutrophil counts occurred; none led to discontinuation. The incidence of malignancies was comparable to that of the general population, and cases were not reminiscent of those observed in immunocompromised patients. Annualised relapse rates remained low, minimal disability progression was observed, with a dose-dependent benefit with teriflunomide 14 mg for several MRI parameters. Conclusion: Teriflunomide had a favourable safety profile for up to 8.5 years. PMID:22307384

Development of consistent hazard controls for DOE transuranic waste operations

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

Woody, W.J.

2007-07-01

This paper describes the results of a re-engineering initiative undertaken with the Department of Energy's (DOE) Office of Environmental Management (EM) in order to standardize hazard analysis assumptions and methods and resulting safety controls applied to multiple transuranic (TRU) waste operations located across the United States. A wide range of safety controls are historically applied to transuranic waste operations, in spite of the fact that these operations have similar operational characteristics and hazard/accident potential. The re-engineering effort supported the development of a DOE technical standard with specific safety controls designated for accidents postulated during waste container retrieval, staging/storage, venting, onsitemore » movements, and characterization activities. Controls cover preventive and mitigative measures; include both hardware and specific administrative controls; and provide protection to the facility worker, onsite co-located workers and the general public located outside of facility boundaries. The Standard development involved participation from all major DOE sites conducting TRU waste operations. Both safety analysts and operations personnel contributed to the re-engineering effort. Acknowledgment is given in particular to the following individuals who formed a core working group: Brenda Hawks, (DOE Oak Ridge Office), Patrice McEahern (CWI-Idaho), Jofu Mishima (Consultant), Louis Restrepo (Omicron), Jay Mullis (DOE-ORO), Mike Hitchler (WSMS), John Menna (WSMS), Jackie East (WSMS), Terry Foppe (CTAC), Carla Mewhinney (WIPP-SNL), Stephie Jennings (WIPP-LANL), Michael Mikolanis (DOESRS), Kraig Wendt (BBWI-Idaho), Lee Roberts (Fluor Hanford), and Jim Blankenhorn (WSRC). Additional acknowledgment is given to Dae Chung (EM) and Ines Triay (EM) for leadership and management of the re-engineering effort. (authors)« less

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.

Anticipatory vigilance: A grounded theory study of minimising risk within the perioperative setting.

PubMed

O'Brien, Brid; Andrews, Tom; Savage, Eileen

2018-01-01

To explore and explain how nurses minimise risk in the perioperative setting. Perioperative nurses care for patients who are having surgery or other invasive explorative procedures. Perioperative care is increasingly focused on how to improve patient safety. Safety and risk management is a global priority for health services in reducing risk. Many studies have explored safety within the healthcare settings. However, little is known about how nurses minimise risk in the perioperative setting. Classic grounded theory. Ethical approval was granted for all aspects of the study. Thirty-seven nurses working in 11 different perioperative settings in Ireland were interviewed and 33 hr of nonparticipant observation was undertaken. Concurrent data collection and analysis was undertaken using theoretical sampling. Constant comparative method, coding and memoing and were used to analyse the data. Participants' main concern was how to minimise risk. Participants resolved this through engaging in anticipatory vigilance (core category). This strategy consisted of orchestrating, routinising and momentary adapting. Understanding the strategies of anticipatory vigilance extends and provides an in-depth explanation of how nurses' behaviour ensures that risk is minimised in a complex high-risk perioperative setting. This is the first theory situated in the perioperative area for nurses. This theory provides a guide and understanding for nurses working in the perioperative setting on how to minimise risk. It makes perioperative nursing visible enabling positive patient outcomes. This research suggests the need for training and education in maintaining safety and minimising risk in the perioperative setting. © 2017 John Wiley & Sons Ltd.

14 CFR Appendix B to Part 415 - Safety Review Document Outline

Code of Federal Regulations, 2013 CFR

2013-01-01

....0Flight Safety (§ 415.115) 4.1Initial Flight Safety Analysis 4.1.1Flight Safety Sub-Analyses, Methods, and... Analysis Data 4.2Radionuclide Data (where applicable) 4.3Flight Safety Plan 4.3.1Flight Safety Personnel 4... Safety (§ 415.117) 5.1Ground Safety Analysis Report 5.2Ground Safety Plan 6.0Launch Plans (§ 415.119 and...

14 CFR Appendix B to Part 415 - Safety Review Document Outline

Code of Federal Regulations, 2014 CFR

2014-01-01

....0Flight Safety (§ 415.115) 4.1Initial Flight Safety Analysis 4.1.1Flight Safety Sub-Analyses, Methods, and... Analysis Data 4.2Radionuclide Data (where applicable) 4.3Flight Safety Plan 4.3.1Flight Safety Personnel 4... Safety (§ 415.117) 5.1Ground Safety Analysis Report 5.2Ground Safety Plan 6.0Launch Plans (§ 415.119 and...

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

NASA Astrophysics Data System (ADS)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. (Abstract shortened by UMI.) (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

Environment, Safety, and Health Self-Assessment Report, Fiscal Year 2008

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

Chernowski, John

2009-02-27

Lawrence Berkeley National Laboratory's Environment, Safety, and Health (ES&H) Self-Assessment Program ensures that Integrated Safety Management (ISM) is implemented institutionally and by all divisions. The Self-Assessment Program, managed by the Office of Contract Assurance (OCA), provides for an internal evaluation of all ES&H programs and systems at LBNL. The functions of the program are to ensure that work is conducted safely, and with minimal negative impact to workers, the public, and the environment. The Self-Assessment Program is also the mechanism used to institute continuous improvements to the Laboratory's ES&H programs. The program is described in LBNL/PUB 5344, Environment, Safety, andmore » Health Self-Assessment Program and is composed of four distinct assessments: the Division Self-Assessment, the Management of Environment, Safety, and Health (MESH) review, ES&H Technical Assurance, and the Appendix B Self-Assessment. The Division Self-Assessment uses the five core functions and seven guiding principles of ISM as the basis of evaluation. Metrics are created to measure performance in fulfilling ISM core functions and guiding principles, as well as promoting compliance with applicable regulations. The five core functions of ISM are as follows: (1) Define the Scope of Work; (2) Identify and Analyze Hazards; (3) Control the Hazards; (4) Perform the Work; and (5) Feedback and Improvement. The seven guiding principles of ISM are as follows: (1) Line Management Responsibility for ES&H; (2) Clear Roles and Responsibilities; (3) Competence Commensurate with Responsibilities; (4) Balanced Priorities; (5) Identification of ES&H Standards and Requirements; (6) Hazard Controls Tailored to the Work Performed; and (7) Operations Authorization. Performance indicators are developed by consensus with OCA, representatives from each division, and Environment, Health, and Safety (EH&S) Division program managers. Line management of each division performs the Division Self-Assessment annually. The primary focus of the review is workplace safety. The MESH review is an evaluation of division management of ES&H in its research and operations, focusing on implementation and effectiveness of the division's ISM plan. It is a peer review performed by members of the LBNL Safety Review Committee (SRC), with staff support from OCA. Each division receives a MESH review every two to four years, depending on the results of the previous review. The ES&H Technical Assurance Program (TAP) provides the framework for systematic reviews of ES&H programs and processes. The intent of ES&H Technical Assurance assessments is to provide assurance that ES&H programs and processes comply with their guiding regulations, are effective, and are properly implemented by LBNL divisions. The Appendix B Performance Evaluation and Measurement Plan (PEMP) requires that LBNL sustain and enhance the effectiveness of integrated safety, health, and environmental protection through a strong and well-deployed system. Information required for Appendix B is provided by EH&S Division functional managers. The annual Appendix B report is submitted at the close of the fiscal year. This assessment is the Department of Energy's (DOE) primary mechanism for evaluating LBNL's contract performance in ISM.« less

Safety testing of GM-rice expressing PHA-E lectin using a new animal test design.

PubMed

Poulsen, Morten; Schrøder, Malene; Wilcks, Andrea; Kroghsbo, Stine; Lindecrona, Rikke Hvid; Miller, Andreas; Frenzel, Thomas; Danier, Jürgen; Rychlik, Michael; Shu, Qingyao; Emami, Kaveh; Taylor, Mark; Gatehouse, Angharad; Engel, Karl-Heinz; Knudsen, Ib

2007-03-01

The 90-day animal study is the core study for the safety assessment of genetically modified foods in the SAFOTEST project. The model compound tested in the 90-day study was a rice variety expressing the kidney bean Phaseolus vulgaris lectin agglutinin E-form (PHA-E lectin). Female Wistar rats were given a nutritionally balanced purified diet with 60% parental rice, 60% PHA-E rice or 60% PHA-E rice spiked with 0.1% recombinant PHA-E lectin for 90 days. This corresponded to a mean daily PHA-E lectin intake of approximately 0, 30 and 100mg/kg body weight for each group, respectively. The spiking was used to increase the specificity and to demonstrate the sensitivity of the study. A range of biological, biochemical, microbiological and pathological parameters were examined and significant differences in weight of small intestine, stomach and pancreas and plasma biochemistry were seen between groups. Included in this paper are also data from the molecular characterisation and chemical analysis of the PHA-E rice, from the construction and production of the PHA-E lectin, and from the preceding 28-day in vivo study where the toxicity of the pure PHA-E lectin was determined. In conclusion, the combined use of information from the compositional analysis, the 28-day study and the characterisation of the PHA-E rice and the PHA-E lectin has improved the design of the 90-day study. The spiking procedure has facilitated the interpretation of the results of the study and transferred it into a valuable tool for the future safety testing of genetically modified foods.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Searching the ASRS Database Using QUORUM Keyword Search, Phrase Search, Phrase Generation, and Phrase Discovery

NASA Technical Reports Server (NTRS)

McGreevy, Michael W.; Connors, Mary M. (Technical Monitor)

2001-01-01

To support Search Requests and Quick Responses at the Aviation Safety Reporting System (ASRS), four new QUORUM methods have been developed: keyword search, phrase search, phrase generation, and phrase discovery. These methods build upon the core QUORUM methods of text analysis, modeling, and relevance-ranking. QUORUM keyword search retrieves ASRS incident narratives that contain one or more user-specified keywords in typical or selected contexts, and ranks the narratives on their relevance to the keywords in context. QUORUM phrase search retrieves narratives that contain one or more user-specified phrases, and ranks the narratives on their relevance to the phrases. QUORUM phrase generation produces a list of phrases from the ASRS database that contain a user-specified word or phrase. QUORUM phrase discovery finds phrases that are related to topics of interest. Phrase generation and phrase discovery are particularly useful for finding query phrases for input to QUORUM phrase search. The presentation of the new QUORUM methods includes: a brief review of the underlying core QUORUM methods; an overview of the new methods; numerous, concrete examples of ASRS database searches using the new methods; discussion of related methods; and, in the appendices, detailed descriptions of the new methods.

C-Mod MHD stability analysis with LHCD

NASA Astrophysics Data System (ADS)

Ebrahimi, Fatima; Bhattacharjee, A.; Delgado, L.; Scott, S.; Wilson, J. R.; Wallace, G. M.; Shiraiwa, S.; Mumgaard, R. T.

2016-10-01

In lower hybrid current drive (LHCD) experiments on the Alcator C-Mod, sawtooth activity could be suppressed as the safety factor q on axis is raised above unity. However, in some of these experiments, after applying LHCD, the onset of MHD mode activity caused the current drive efficiency to significantly drop. Here, we study the stability of these experiments by performing MHD simulations using the NIMROD code starting with experimental EFIT equilibria. First, consistent with the LHCD experiment with no signature of MHD activity, MHD mode activity was also absent in the simulations. Second, for experiments with MHD mode activity, we find that a core n=1 reconnecting mode with dominate poloidal modes of m=2,3 is unstable. This mode is a resistive current-driven mode as its growth rate scales with a negative power of the Lundquist number in the simulations. In addition, with further enhanced reversed-shear q profile in the simulations, a core double tearing mode is found to be unstable. This work is supported by U.S. DOE cooperative agreement DE-FC02-99ER54512 using the Alcator C-Mod tokamak, a DOE Office of Science user facility.

Standardized Curriculum for General Industrial Maintenance Trades.

ERIC Educational Resources Information Center

Mississippi State Dept. of Education, Jackson. Office of Vocational, Technical and Adult Education.

Standardized vocational education course titles and core contents for two courses in Mississippi are provided: general industrial maintenance trades I and II. The first course contains the following units: (1) orientation and leadership activities; (2) safety; (3) blueprint reading; (4) oxyacetylene cutting; (5) preventative maintenance; (6)…

Health: Overview. Interim Guide.

ERIC Educational Resources Information Center

Manitoba Dept. of Education, Winnipeg.

The Manitoba Health Education Curriculum focuses on promoting the development of positive lifestyle practices in students from kindergarten through grade nine. The core units of the program are: (1) social-emotional well-being; (2) physical well-being; (3) nutrition; (4) dental health; (5) safety; and (6) environmental health. This overview offers…

Introduction to Horticulture. Teacher Edition. Horticulture Series.

ERIC Educational Resources Information Center

Oklahoma State Dept. of Vocational and Technical Education, Stillwater. Curriculum and Instructional Materials Center.

This publication is designed to provide a core of instruction for the many different fields in agricultural/horticultural education. This course contains 21 instructional units that cover the following topics: introduction to horticulture; beginning a career in horticulture; hand and power tools; introduction to safety; growing facilities;…

30 CFR 57.4760 - Shaft mines.

Code of Federal Regulations, 2014 CFR

2014-07-01

..., but without an insulation core, are acceptable if an automatic sprinkler or deluge system is installed... HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Fire Prevention and Control... following means to control the spread of fire, smoke, and toxic gases underground in the event of a fire...

Standardized Curriculum for Electricity/Electronics.

ERIC Educational Resources Information Center

Mississippi State Dept. of Education, Jackson. Office of Vocational, Technical and Adult Education.

Standardized vocational education course titles and core contents are provided for two courses in Mississippi: electricity/electronics I and II. The first course contains the following units: (1) orientation, safety, and leadership; (2) basic principles of electricity/electronics; (3) direct current (DC) theory; (4) magnetism and DC motors; (5)…

CERA: Clerkships Need National Curricula on Care Delivery, Awareness of Their NCC Gaps.

PubMed

Cochella, Susan; Liaw, Winston; Binienda, Juliann; Hustedde, Carol

2016-06-01

The Society of Teachers of Family Medicine's (STFM) National Clerkship Curriculum (NCC) was created to standardize and improve teaching of a minimum core curriculum in family medicine clerkships, promoting the Triple Aim of better care and population health at lower cost. It includes competencies all clerkships should teach and tools to support clerkship directors (CDs). This 2014 CERA survey of clerkship directors is one of several needs assessments that guide STFM's NCC Editorial Board in targeting improvements and peer-review processes. CERA's 2014 survey of CDs was sent to all 137 CDs at US and Canadian allopathic medical schools. Primary aims included: (1) Identify curricular topics of greatest need, (2) Inventory the percent of family medicine clerkships teaching each NCC topic, and (3) Determine if longitudinal or blended clerkship have unique needs. This survey also assessed use of NCC to advocate for teaching resources and collaborate with colleagues at other institutions. Ninety-one percent of CDs completed the survey. Sixty-four percent reported their clerkship covers all of the NCC minimum core, but on detailed analysis, only 1% teach all topics. CDs need curricula on care delivery topics (cost-effective approach to acute care, role of family medicine in the health care system, quality/safety, and comorbid substance abuse). Single-question assessments overestimate the percentage of clerkships teaching all of the NCC minimum core. Clerkships need national curricula on care delivery topics and tools to help them find their curricular gaps.

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

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

Shemon, Emily R.

2016-10-10

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling andmore » simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.« less

Publications - GMC 418 | Alaska Division of Geological & Geophysical

Science.gov Websites

DGGS GMC 418 Publication Details Title: Porosity, permeability, grain density core analysis (CT scans , permeability, grain density core analysis (CT scans), and core photos from the ConocoPhillips N. Cook Inlet

Patient safety competence for final-year health professional students: Perceptions of effectiveness of an interprofessional education course.

PubMed

Hwang, Jee-In; Yoon, Tai-Young; Jin, Hyeon-Jeong; Park, Yikyun; Park, Ju-Young; Lee, Beom-Joon

2016-11-01

As final-year medical and nursing students will soon play key roles in frontline patient care, their preparedness for safe, reliable care provision is of special importance. We assessed patient safety competencies of final-year health profession students, and the effect of a 1-day patient safety education programme on these competencies. A cross-sectional survey was conducted with 233 students in three colleges of medicine, nursing, and traditional medicine in Seoul. A before-and-after study followed to evaluate the effectiveness of the curriculum. Patient safety competency was measured using the Health-Professional Education for Patients Safety Survey (H-PEPSS) and an objective patient safety knowledge test. The mean scores were 3.4 and 1.7 out of 5.0, respectively. The communication domain was rated the highest and the teamwork domain was rated the lowest. H-PEPSS scores significantly differed between the students from three colleges. The 1-day patient safety education curriculum significantly improved H-PEPSS and knowledge test scores. These results indicated that strengthening patient safety competencies, especially teamwork, of students is required in undergraduate healthcare curricula. A 1-day interprofessional patient safety education programme may be a promising strategy. The findings suggest that interprofessional patient safety education needs to be implemented as a core undergraduate course to improve students' safety competence.

75 FR 17604 - Federal Motor Vehicle Safety Standards; Roof Crush Resistance

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-07

... Safety Analysis & Forensic Engineering, LLC (SAFE) brought to our attention errors in the preamble that incorrectly attributed to it the comments of another organization, Safety Analysis, Inc. Both of these... Safety Analysis, Inc. SAFE noted that there is no affiliation between SAFE and Safety Analysis, Inc. and...

Global Precipitation Measurement (GPM) Mission

NASA Image and Video Library

2014-02-22

NASA GPM Safety Quality and Assurance, Shirley Dion, and, NASA GPM Quality and Assurance, Larry Morgan, monitor the all-day launch simulation for the Global Precipitation Measurement (GPM) Core Observatory at the Spacecraft Test and Assembly Building 2 (STA2), Saturday, Feb. 22, 2014, Tanegashima Space Center (TNSC), Tanegashima Island, Japan. Japan Aerospace Exploration Agency (JAXA) plans to launch an H-IIA rocket carrying the GPM Core Observatory on Feb. 28, 2014. The NASA-JAXA GPM spacecraft will collect information that unifies data from an international network of existing and future satellites to map global rainfall and snowfall every three hours. Photo Credit: (NASA/Bill Ingalls)

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

The influence of geometric imperfections on the stability of three-layer beams with foam core

NASA Astrophysics Data System (ADS)

Wstawska, Iwona

2017-01-01

The main objective of this work is the numerical analysis (FE analysis) of stability of three-layer beams with metal foam core (alumina foam core). The beams were subjected to pure bending. The analysis of the local buckling was performed. Furthermore, the influence of geometric parameters of the beam and material properties of the core (linear and non-linear model) on critical loads values and buckling shape were also investigated. The calculations were made on a family of beams with different mechanical properties of the core (elastic and elastic-plastic material). In addition, the influence of geometric imperfections on deflection and normal stress values of the core and the faces has been evaluated.

Team-Based Care: A Concept Analysis.

PubMed

Baik, Dawon

2017-10-01

The purpose of this concept analysis is to clarify and analyze the concept of team-based care in clinical practice. Team-based care has garnered attention as a way to enhance healthcare delivery and patient care related to quality and safety. However, there is no consensus on the concept of team-based care; as a result, the lack of common definition impedes further studies on team-based care. This analysis was conducted using Walker and Avant's strategy. Literature searches were conducted using PubMed, Cumulative Index to Nursing and Allied Health Literature (CINAHL), and PsycINFO, with a timeline from January 1985 to December 2015. The analysis demonstrates that the concept of team-based care has three core attributes: (a) interprofessional collaboration, (b) patient-centered approach, and (c) integrated care process. This is accomplished through understanding other team members' roles and responsibilities, a climate of mutual respect, and organizational support. Consequences of team-based care are identified with three aspects: (a) patient, (b) healthcare professional, and (c) healthcare organization. This concept analysis helps better understand the characteristics of team-based care in the clinical practice as well as promote the development of a theoretical definition of team-based care. © 2016 Wiley Periodicals, Inc.

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

DOE PAGES

Davidson, Eva E.; Betzler, Benjamin R.; Chandler, David; ...

2017-08-01

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed andmore » benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.« less

Trauma-Informed Social Policy: A Conceptual Framework for Policy Analysis and Advocacy

PubMed Central

Murshid, Nadine Shaanta

2016-01-01

Trauma-informed care is a service provision model used across a range of practice settings. Drawing on an extensive body of research on trauma (broadly defined as experiences that produce enduring emotional pain and distress) and health outcomes, we have argued that the principles of trauma-informed care can be extended to social policy. Citing a variety of health-related policy examples, we have described how policy can better reflect 6 core principles of trauma-informed care: safety, trustworthiness and transparency, collaboration, empowerment, choice, and intersectionality. This framework conveys a politicized understanding of trauma, reflecting the reality that trauma and its effects are not equally distributed, and offers a pathway for public health professionals to disrupt trauma-driven health disparities through policy action. PMID:26691122

14 CFR 415.115 - Flight safety.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 14 Aeronautics and Space 4 2012-01-01 2012-01-01 false Flight safety. 415.115 Section 415.115... From a Non-Federal Launch Site § 415.115 Flight safety. (a) Flight safety analysis. An applicant's safety review document must describe each analysis method employed to meet the flight safety analysis...

14 CFR 415.115 - Flight safety.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Flight safety. 415.115 Section 415.115... From a Non-Federal Launch Site § 415.115 Flight safety. (a) Flight safety analysis. An applicant's safety review document must describe each analysis method employed to meet the flight safety analysis...

14 CFR 415.115 - Flight safety.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 14 Aeronautics and Space 4 2013-01-01 2013-01-01 false Flight safety. 415.115 Section 415.115... From a Non-Federal Launch Site § 415.115 Flight safety. (a) Flight safety analysis. An applicant's safety review document must describe each analysis method employed to meet the flight safety analysis...

14 CFR 415.115 - Flight safety.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 14 Aeronautics and Space 4 2011-01-01 2011-01-01 false Flight safety. 415.115 Section 415.115... From a Non-Federal Launch Site § 415.115 Flight safety. (a) Flight safety analysis. An applicant's safety review document must describe each analysis method employed to meet the flight safety analysis...

14 CFR 415.115 - Flight safety.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 14 Aeronautics and Space 4 2014-01-01 2014-01-01 false Flight safety. 415.115 Section 415.115... From a Non-Federal Launch Site § 415.115 Flight safety. (a) Flight safety analysis. An applicant's safety review document must describe each analysis method employed to meet the flight safety analysis...

A Finite Element Analysis for Predicting the Residual Compressive Strength of Impact-Damaged Sandwich Panels

NASA Technical Reports Server (NTRS)

Ratcliffe, James G.; Jackson, Wade C.

2008-01-01

A simple analysis method has been developed for predicting the residual compressive strength of impact-damaged sandwich panels. The method is tailored for honeycomb core-based sandwich specimens that exhibit an indentation growth failure mode under axial compressive loading, which is driven largely by the crushing behavior of the core material. The analysis method is in the form of a finite element model, where the impact-damaged facesheet is represented using shell elements and the core material is represented using spring elements, aligned in the thickness direction of the core. The nonlinear crush response of the core material used in the analysis is based on data from flatwise compression tests. A comparison with a previous analysis method and some experimental data shows good agreement with results from this new approach.

A Finite Element Analysis for Predicting the Residual Compression Strength of Impact-Damaged Sandwich Panels

NASA Technical Reports Server (NTRS)

Ratcliffe, James G.; Jackson, Wade C.

2008-01-01

A simple analysis method has been developed for predicting the residual compression strength of impact-damaged sandwich panels. The method is tailored for honeycomb core-based sandwich specimens that exhibit an indentation growth failure mode under axial compression loading, which is driven largely by the crushing behavior of the core material. The analysis method is in the form of a finite element model, where the impact-damaged facesheet is represented using shell elements and the core material is represented using spring elements, aligned in the thickness direction of the core. The nonlinear crush response of the core material used in the analysis is based on data from flatwise compression tests. A comparison with a previous analysis method and some experimental data shows good agreement with results from this new approach.

HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

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

John D. Bess; Barbara H. Dolphin; James W. Sterbentz

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.« less

HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

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

John D. Bess; Barbara H. Dolphin; James W. Sterbentz

2012-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.« less

Post-market drug evaluation research training capacity in Canada: an environmental scan of Canadian educational institutions.

PubMed

Wiens, Matthew O; Soon, Judith A; MacLeod, Stuart M; Sharma, Sunaina; Patel, Anik

2014-01-01

Ongoing efforts by Health Canada intended to modernize the legislation and regulation of pharmaceuticals will help improve the safety and effectiveness of drug products. It will be imperative to ensure that comprehensive and specialized training sites are available to train researchers to support the regulation of therapeutic products. The objective of this educational institution inventory was to conduct an environmental scan of educational institutions in Canada able to train students in areas of post-market drug evaluation research. A systematic web-based environmental scan of Canadian institutions was conducted. The website of each university was examined for potential academic programs. Six core programmatic areas were determined a priori as necessary to train competent post-market drug evaluation researchers. These included biostatistics, epidemiology, pharmacoepidemiology, health economics or pharmacoeconomics, pharmacogenetics or pharmacogenomics and patient safety/pharmacovigilance. Twenty-three academic institutions were identified that had the potential to train students in post-market drug evaluation research. Overall, 23 institutions taught courses in epidemiology, 22 in biostatistics, 17 in health economics/pharmacoeconomics, 5 in pharmacoepidemiology, 5 in pharmacogenetics/pharmacogenomics, and 3 in patient safety/pharmacovigilance. Of the 23 institutions, only the University of Ottawa offered six core courses. Two institutions offered five, seven offered four and the remaining 14 offered three or fewer. It is clear that some institutions may offer programs not entirely reflected in the nomenclature used for this review. As Heath Canada moves towards a more progressive licensing framework, augmented training to increase research capacity and expertise in drug safety and effectiveness is timely and necessary.

Brave New Workplace: Technology and Work in the New Economy.

ERIC Educational Resources Information Center

Wallace, Michael

1989-01-01

Technological innovations in factories and offices are examined in terms of 10 core issues: "high flex" workplace; control of work; organizational change; impact on skill; unemployment; educational needs and retraining; changing occupational structures; safety and health; interaction of work, leisure, and family; and quality of working life. The…

Curriculum Guide for Electronics in Technology Education.

ERIC Educational Resources Information Center

Connecticut Industrial Technology Association.

Consistent with the principles of the Connecticut Common Core of Learning, this competency-based curriculum guide for electronics provides a reference guide for educators to research and prepare for teaching the field of electronics. The guide contains 22 units that cover the following topics: theory of matter; safety; direct current; magnetism;…

The Digital School Library: A World-Wide Development and a Fascinating Challenge.

ERIC Educational Resources Information Center

Loertscher, David

2003-01-01

Explores the academic environment of a total information system for school libraries based on the idea of a digital intranet. Discusses safety; customization; the core library collection; curriculum-specific collections; access to short-term resources; Internet access; personalized features; search engines; equity issues; and staffing. (LRW)

Standardized Curriculum for Heating and Air Conditioning.

ERIC Educational Resources Information Center

Mississippi State Dept. of Education, Jackson. Office of Vocational, Technical and Adult Education.

Standardized vocational education course titles and core contents for two courses in Mississippi are provided: heating and air conditioning I and II. The first course contains the following units: (1) orientation; (2) safety; (3) refrigeration gauges and charging cylinder; (4) vacuum pump service operations; (5) locating refrigerant leaks; (6)…

Allied Health Occupations II (Health Careers--Core Curriculum).

ERIC Educational Resources Information Center

Middletown Public Schools, CT.

This volume outlines the requirements and content of a second-year course in allied health occupations education that is designed to provide students with background informational material and practical skills used in various health fields. Addressed in the individual units of the course are the following topics: safety; ethical and legal…

Standardized Curriculum for Food Production, Management and Services.

ERIC Educational Resources Information Center

Mississippi State Dept. of Education, Jackson. Office of Vocational, Technical and Adult Education.

Standardized vocational education course titles and core contents for two courses in Mississippi are provided: food production, management, and services I and II. The first course contains the following units: (1) Vocational Industrial Clubs of America (VICA); (2) sanitation; (3) safety; (4) front of the house operations; (5) beverages; (6) food…

100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

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

HARRINGTON RA

2010-01-15

On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, tomore » all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery. In brief, the Path Forward was developed to reconsider potential open air demolition areas; characterize to determine if any zircaloy exists, evaluate existing concrete data to determine additional characterization needs, size the new building to accommodate human machine interface and tooling, consider bucket thumb and use ofshape-charges in design, and finally to utilize complex-wide and industry explosive demolition lessons learned in the design approach. Appendix B documents these results from the team's use ofValue Engineering process tools entitled Weighted Analysis Alternative Matrix, Matrix Conclusions, Evaluation Criteria, and Alternative Advantages and Disadvantages. These results were further supported with the team's validation of parking-lot information sheets: memories (potential ideas to consider), issues/concerns, and assumptions, contained in Appendix C. Appendix C also includes the recorded workshop flipchart notes taken from the SAR Alternatives and Project Overview presentations. The SAR workshop presentations, including a 3-D graphic illustration demonstration video have been retained in the CHPRC project file, and were not included in this report due to size limitations. The workshop concluded with a round robin close-out where each member was engaged for any last minute items and meeting utility. In summary, the team felt the session was value added and looked forward to proceeding with the recommended actions and conceptual design.« less

Drug-resistant tuberculosis clinical trials: proposed core research definitions in adults.

PubMed

Furin, J; Alirol, E; Allen, E; Fielding, K; Merle, C; Abubakar, I; Andersen, J; Davies, G; Dheda, K; Diacon, A; Dooley, K E; Dravnice, G; Eisenach, K; Everitt, D; Ferstenberg, D; Goolam-Mahomed, A; Grobusch, M P; Gupta, R; Harausz, E; Harrington, M; Horsburgh, C R; Lienhardt, C; McNeeley, D; Mitnick, C D; Nachman, S; Nahid, P; Nunn, A J; Phillips, P; Rodriguez, C; Shah, S; Wells, C; Thomas-Nyang'wa, B; du Cros, P

2016-03-01

Drug-resistant tuberculosis (DR-TB) is a growing public health problem, and for the first time in decades, new drugs for the treatment of this disease have been developed. These new drugs have prompted strengthened efforts in DR-TB clinical trials research, and there are now multiple ongoing and planned DR-TB clinical trials. To facilitate comparability and maximise policy impact, a common set of core research definitions is needed, and this paper presents a core set of efficacy and safety definitions as well as other important considerations in DR-TB clinical trials work. To elaborate these definitions, a search of clinical trials registries, published manuscripts and conference proceedings was undertaken to identify groups conducting trials of new regimens for the treatment of DR-TB. Individuals from these groups developed the core set of definitions presented here. Further work is needed to validate and assess the utility of these definitions but they represent an important first step to ensure there is comparability in clinical trials on multidrug-resistant TB.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Modeling of grain-oriented Si-steel and amorphous alloy iron core under ferroresonance using Jiles-Atherton hysteresis method

NASA Astrophysics Data System (ADS)

Sima, Wenxia; Zou, Mi; Yang, Ming; Yang, Qing; Peng, Daixiao

2018-05-01

Amorphous alloy is increasingly widely used in the iron core of power transformer due to its excellent low loss performance. However, its potential harm to the power system is not fully studied during the electromagnetic transients of the transformer. This study develops a simulation model to analyze the effect of transformer iron core materials on ferroresonance. The model is based on the transformer π equivalent circuit. The flux linkage-current (ψ-i) Jiles-Atherton reactor is developed in an Electromagnetic Transients Program-Alternative Transients Program and is used to represent the magnetizing branches of the transformer model. Two ferroresonance cases are studied to compare the performance of grain-oriented Si-steel and amorphous alloy cores. The ferroresonance overvoltage and overcurrent are discussed under different system parameters. Results show that amorphous alloy transformer generates higher voltage and current than those of grain-oriented Si-steel transformer and significantly harms the power system safety.

Evaluation of Analysis Techniques for Fluted-Core Sandwich Cylinders

NASA Technical Reports Server (NTRS)

Lovejoy, Andrew E.; Schultz, Marc R.

2012-01-01

Buckling-critical launch-vehicle structures require structural concepts that have high bending stiffness and low mass. Fluted-core, also known as truss-core, sandwich construction is one such concept. In an effort to identify an analysis method appropriate for the preliminary design of fluted-core cylinders, the current paper presents and compares results from several analysis techniques applied to a specific composite fluted-core test article. The analysis techniques are evaluated in terms of their ease of use and for their appropriateness at certain stages throughout a design analysis cycle (DAC). Current analysis techniques that provide accurate determination of the global buckling load are not readily applicable early in the DAC, such as during preliminary design, because they are too costly to run. An analytical approach that neglects transverse-shear deformation is easily applied during preliminary design, but the lack of transverse-shear deformation results in global buckling load predictions that are significantly higher than those from more detailed analysis methods. The current state of the art is either too complex to be applied for preliminary design, or is incapable of the accuracy required to determine global buckling loads for fluted-core cylinders. Therefore, it is necessary to develop an analytical method for calculating global buckling loads of fluted-core cylinders that includes transverse-shear deformations, and that can be easily incorporated in preliminary design.

Occupational health and safety-ergonomics improvement as a corporate responsibility of a Bali handicraft company: a case study.

PubMed

Purnawati, Susy

2007-12-01

The issue of corporate social responsibility is nowadays becoming popular around industrial communities. The support for the issue has initially spread since the adoption in 1998 of the ILO Declaration concerning fundamental principles and rights at work and then followed up by industries in developed countries. A case study was done from February to August 2006 at a handicraft company in Bali in order to find out the core application of the issue at the enterprise level. The study was conducted by observation in the field of the factory and suppliers, taking photos and interviewing management and employees of the company. The results of the study show that the company has already executed the activities that reflect the application of the core principles. The activities included programs which concerned not only the business corporate community but also wider communities. With regard to the business corporate community, the company had improved the conditions related to ergonomics and occupational health and safety. The improvement was done by referring to the external audit. At the national community level, the company had participated in the recovery measures of national disasters by helping small industries revive. It is hoped that this core program is soon copied by other companies considering that it is very beneficial to the communities and companies.

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

s-core network decomposition: A generalization of k-core analysis to weighted networks

NASA Astrophysics Data System (ADS)

Eidsaa, Marius; Almaas, Eivind

2013-12-01

A broad range of systems spanning biology, technology, and social phenomena may be represented and analyzed as complex networks. Recent studies of such networks using k-core decomposition have uncovered groups of nodes that play important roles. Here, we present s-core analysis, a generalization of k-core (or k-shell) analysis to complex networks where the links have different strengths or weights. We demonstrate the s-core decomposition approach on two random networks (ER and configuration model with scale-free degree distribution) where the link weights are (i) random, (ii) correlated, and (iii) anticorrelated with the node degrees. Finally, we apply the s-core decomposition approach to the protein-interaction network of the yeast Saccharomyces cerevisiae in the context of two gene-expression experiments: oxidative stress in response to cumene hydroperoxide (CHP), and fermentation stress response (FSR). We find that the innermost s-cores are (i) different from innermost k-cores, (ii) different for the two stress conditions CHP and FSR, and (iii) enriched with proteins whose biological functions give insight into how yeast manages these specific stresses.

ROMUSE 2.0 User Manual

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

Khuwaileh, Bassam; Turinsky, Paul; Williams, Brian J.

2016-10-04

ROMUSE (Reduced Order Modeling Based Uncertainty/Sensitivity Estimator) is an effort within the Consortium for Advanced Simulation of Light water reactors (CASL) to provide an analysis tool to be used in conjunction with reactor core simulators, especially the Virtual Environment for Reactor Applications (VERA). ROMUSE is written in C++ and is currently capable of performing various types of parameters perturbations, uncertainty quantification, surrogate models construction and subspace analysis. Version 2.0 has the capability to interface with DAKOTA which gives ROMUSE access to the various algorithms implemented within DAKOTA. ROMUSE is mainly designed to interface with VERA and the Comprehensive Modeling andmore » Simulation Suite for Nuclear Safety Analysis and Design (SCALE) [1,2,3], however, ROMUSE can interface with any general model (e.g. python and matlab) with Input/Output (I/O) format that follows the Hierarchical Data Format 5 (HDF5). In this brief user manual, the use of ROMUSE will be overviewed and example problems will be presented and briefly discussed. The algorithms provided here range from algorithms inspired by those discussed in Ref.[4] to nuclear-specific algorithms discussed in Ref. [3].« less

14 CFR 417.221 - Time delay analysis.

Code of Federal Regulations, 2010 CFR

2010-01-01

... OF TRANSPORTATION LICENSING LAUNCH SAFETY Flight Safety Analysis § 417.221 Time delay analysis. (a) General. A flight safety analysis must include a time delay analysis that establishes the mean elapsed time between the violation of a flight termination rule and the time when the flight safety system is...

14 CFR 417.221 - Time delay analysis.

Code of Federal Regulations, 2011 CFR

2011-01-01

... OF TRANSPORTATION LICENSING LAUNCH SAFETY Flight Safety Analysis § 417.221 Time delay analysis. (a) General. A flight safety analysis must include a time delay analysis that establishes the mean elapsed time between the violation of a flight termination rule and the time when the flight safety system is...

Single Event Effects (SEE) Testing of Embedded DSP Cores within Microsemi RTAX4000D Field Programmable Gate Array (FPGA) Devices

NASA Technical Reports Server (NTRS)

Perez, Christopher E.; Berg, Melanie D.; Friendlich, Mark R.

2011-01-01

Motivation for this work is: (1) Accurately characterize digital signal processor (DSP) core single-event effect (SEE) behavior (2) Test DSP cores across a large frequency range and across various input conditions (3) Isolate SEE analysis to DSP cores alone (4) Interpret SEE analysis in terms of single-event upsets (SEUs) and single-event transients (SETs) (5) Provide flight missions with accurate estimate of DSP core error rates and error signatures.

Occupational safety and HIV risk among female sex workers in China: A mixed-methods analysis of sex-work harms and mommies

PubMed Central

Yi, Huso; Zheng, Tiantian; Wan, Yanhai; Mantell, Joanne E.; Park, Minah; Csete, Joanne

2013-01-01

Female sex workers (FSWs) in China are exposed to multiple work-related harms that increase HIV vulnerability. Using mixed-methods, we explored the social-ecological aspects of sexual risk among 348 FSWs in Beijing. Sex-work harms were assessed by property stolen, being underpaid or not paid at all, verbal and sexual abuse, forced drinking; and forced sex more than once. The majority (90%) reported at least one type of harm, 38% received harm protection from ‘mommies’ (i.e., managers) and 32% reported unprotected sex with clients. In multivariate models, unprotected sex was significantly associated with longer involvement in sex work, greater exposure to harms, and no protection from mommies. Mommies’ protection moderated the effect of sex-work harms on unprotected sex with clients. Our ethnography indicated that mommies played a core role in sex-work networks. Such networks provide a basis for social capital; they are not only profitable economically, but also protect FSWs from sex-work harms. Effective HIV prevention interventions for FSWs in China must address the occupational safety and health of FSWs by facilitating social capital and protection agency (e.g., mommies) in the sex-work industry. PMID:22375698

An improved method for field extraction and laboratory analysis of large, intact soil cores

USGS Publications Warehouse

Tindall, J.A.; Hemmen, K.; Dowd, J.F.

1992-01-01

Various methods have been proposed for the extraction of large, undisturbed soil cores and for subsequent analysis of fluid movement within the cores. The major problems associated with these methods are expense, cumbersome field extraction, and inadequate simulation of unsaturated flow conditions. A field and laboratory procedure is presented that is economical, convenient, and simulates unsaturated and saturated flow without interface flow problems and can be used on a variety of soil types. In the field, a stainless steel core barrel is hydraulically pressed into the soil (30-cm diam. and 38 cm high), the barrel and core are extracted from the soil, and after the barrel is removed from the core, the core is then wrapped securely with flexible sheet metal and a stainless mesh screen is attached to the bottom of the core for support. In the laboratory the soil core is set atop a porous ceramic plate over which a soil-diatomaceous earth slurry has been poured to assure good contact between plate and core. A cardboard cylinder (mold) is fastened around the core and the empty space filled with paraffin wax. Soil cores were tested under saturated and unsaturated conditions using a hanging water column for potentials ???0. Breakthrough curves indicated that no interface flow occurred along the edge of the core. This procedure proved to be reliable for field extraction of large, intact soil cores and for laboratory analysis of solute transport.

Transportation systems safety hazard analysis tool (SafetyHAT) user guide (version 1.0)

DOT National Transportation Integrated Search

2014-03-24

This is a user guide for the transportation system Safety Hazard Analysis Tool (SafetyHAT) Version 1.0. SafetyHAT is a software tool that facilitates System Theoretic Process Analysis (STPA.) This user guide provides instructions on how to download, ...

10 CFR 830.206 - Preliminary documented safety analysis.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Preliminary documented safety analysis. 830.206 Section 830.206 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.206 Preliminary documented safety analysis. If construction begins after December 11, 2000, the contractor...

Sensor Failure Detection of FASSIP System using Principal Component Analysis

NASA Astrophysics Data System (ADS)

Sudarno; Juarsa, Mulya; Santosa, Kussigit; Deswandri; Sunaryo, Geni Rina

2018-02-01

In the nuclear reactor accident of Fukushima Daiichi in Japan, the damages of core and pressure vessel were caused by the failure of its active cooling system (diesel generator was inundated by tsunami). Thus researches on passive cooling system for Nuclear Power Plant are performed to improve the safety aspects of nuclear reactors. The FASSIP system (Passive System Simulation Facility) is an installation used to study the characteristics of passive cooling systems at nuclear power plants. The accuracy of sensor measurement of FASSIP system is essential, because as the basis for determining the characteristics of a passive cooling system. In this research, a sensor failure detection method for FASSIP system is developed, so the indication of sensor failures can be detected early. The method used is Principal Component Analysis (PCA) to reduce the dimension of the sensor, with the Squarred Prediction Error (SPE) and statistic Hotteling criteria for detecting sensor failure indication. The results shows that PCA method is capable to detect the occurrence of a failure at any sensor.

Structural Design of Ares V Interstage Composite Structure

NASA Technical Reports Server (NTRS)

Sleigh, David W.; Sreekantamurthy, Thammaiah; Kosareo, Daniel N.; Martin, Robert A.; Johnson, Theodore F.

2011-01-01

Preliminary and detailed design studies were performed to mature composite structural design concepts for the Ares V Interstage structure as a part of NASA s Advanced Composite Technologies Project. Aluminum honeycomb sandwich and hat-stiffened composite panel structural concepts were considered. The structural design and analysis studies were performed using HyperSizer design sizing software and MSC Nastran finite element analysis software. System-level design trade studies were carried out to predict weight and margins of safety for composite honeycomb-core sandwich and composite hat-stiffened skin design concepts. Details of both preliminary and detailed design studies are presented in the paper. For the range of loads and geometry considered in this work, the hat-stiffened designs were found to be approximately 11-16 percent lighter than the sandwich designs. A down-select process was used to choose the most favorable structural concept based on a set of figures of merit, and the honeycomb sandwich design was selected as the best concept based on advantages in manufacturing cost.

A streamlined failure mode and effects analysis

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

Ford, Eric C., E-mail: eford@uw.edu; Smith, Koren; Terezakis, Stephanie

Purpose: Explore the feasibility and impact of a streamlined failure mode and effects analysis (FMEA) using a structured process that is designed to minimize staff effort. Methods: FMEA for the external beam process was conducted at an affiliate radiation oncology center that treats approximately 60 patients per day. A structured FMEA process was developed which included clearly defined roles and goals for each phase. A core group of seven people was identified and a facilitator was chosen to lead the effort. Failure modes were identified and scored according to the FMEA formalism. A risk priority number,RPN, was calculated and usedmore » to rank failure modes. Failure modes with RPN > 150 received safety improvement interventions. Staff effort was carefully tracked throughout the project. Results: Fifty-two failure modes were identified, 22 collected during meetings, and 30 from take-home worksheets. The four top-ranked failure modes were: delay in film check, missing pacemaker protocol/consent, critical structures not contoured, and pregnant patient simulated without the team's knowledge of the pregnancy. These four failure modes hadRPN > 150 and received safety interventions. The FMEA was completed in one month in four 1-h meetings. A total of 55 staff hours were required and, additionally, 20 h by the facilitator. Conclusions: Streamlined FMEA provides a means of accomplishing a relatively large-scale analysis with modest effort. One potential value of FMEA is that it potentially provides a means of measuring the impact of quality improvement efforts through a reduction in risk scores. Future study of this possibility is needed.« less

Strengthening core public health capacity based on the implementation of the International Health Regulations (IHR) (2005): Chinese lessons

PubMed Central

Liu, Bin; Sun, Yan; Dong, Qian; Zhang, Zongjiu; Zhang, Liang

2015-01-01

As an international legal instrument, the International Health Regulations (IHR) is internationally binding in 196 countries, especially in all the member states of the World Health Organization (WHO). The IHR aims to prevent, protect against, control, and respond to the international spread of disease and aims to cut out unnecessary interruptions to traffic and trade. To meet IHR requirements, countries need to improve capacity construction by developing, strengthening, and maintaining core response capacities for public health risk and Public Health Emergency of International Concern (PHEIC). In addition, all the related core capacity requirements should be met before June 15, 2012. If not, then the deadline can be extended until 2016 upon request by countries. China has promoted the implementation of the IHR comprehensively, continuingly strengthening the core public health capacity and advancing in core public health emergency capacity building, points of entry capacity building, as well as risk prevention and control of biological events (infectious diseases, zoonotic diseases, and food safety), radiological, nuclear, and chemical events, and other catastrophic events. With significant progress in core capacity building, China has dealt with many public health emergencies successfully, ensuring that its core public health capacity has met the IHR requirements, which was reported to WHO in June 2014. This article describes the steps, measures, and related experiences in the implementation of IHR in China. PMID:26029897

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

Low density bismaleimide-carbon microballoon composites. [aircraft and submarine compartment safety

NASA Technical Reports Server (NTRS)

Kourtides, D. A.; Parker, J. A. (Inventor)

1978-01-01

A process is described for constructing for a composite laminate structure which exhibits a high resistance to heat and flame provides safer interior structures for aircraft and submarine compartments. Composite laminate structures are prepared by the bismaleimide resin preimpregnation of a fiberglass cloth to form a face sheet which is bonded with a bismaleimide hot melt adhesive to a porous core structure selected from the group consisting of polyamide paper and bismaleimide-glass fabric which is filled with carbon microballoons. The carbon microballoons are prepared by pyrolyzing phenolic micro-balloons in the presence of nitrogen. A slurry of the carbon microballoons is prepared to fill the porous core structure. The porous core structure and face sheet are bonded to provide panel structures exhibiting increased mechanical capacities and lower oxygen limit values and smoke density values.

Survey on patient safety climate in public hospitals in China.

PubMed

Zhou, Ping; Bundorf, M Kate; Gu, Jianjun; He, Xiaoyan; Xue, Di

2015-02-07

Patient safety climate has been recognized as a core determinant for improving safety in hospitals. Describing workforce perceptions of patient safety climate is an important part of safety climate management. This study aimed to describe staff's perceptions of patient safety climate in public hospitals in Shanghai, China and to determine how perceptions of patient safety climate differ between different types of workers in the U.S. and China. Survey of employees of 6 secondary, general public hospitals in Shanghai conducted during 2013 using a modified version of the U.S. Patient Safety Climate in Health Care Organizations (PSCHO) tool. The percentage of "problematic responses" (PPRs) was used to measure safety climate, and the PPRs were compared among employees with different job types, using χ (2) tests and multivariate regression models. Perceptions of patient safety climate were relatively positive among hospital employees and similar to those of employees in U.S. hospitals along most dimensions. For workers in Chinese hospitals, the scales of "fear of blame" and "fear of shame" had the highest PPRs, whereas in the United States the scale of "fear of shame" had among the lowest PPRs. As in the United States, hospital managers in China perceived a more positive patient safety climate overall than other types of personnel. "Fear of shame" and "fear of blame" may be important barriers to improvement of patient safety in Chinese hospitals. Research on the effect of patient safety climate on outcomes is necessary to implement effective polices to improve patient safety and quality outcomes in China.

How do the top 12 pharmaceutical companies operate safety pharmacology?

PubMed

Ewart, Lorna; Gallacher, David J; Gintant, Gary; Guillon, Jean-Michel; Leishman, Derek; Levesque, Paul; McMahon, Nick; Mylecraine, Lou; Sanders, Martin; Suter, Willi; Wallis, Rob; Valentin, Jean-Pierre

2012-09-01

How does safety pharmacology operate in large pharmaceutical companies today? By understanding our current position, can we prepare safety pharmacology to successfully navigate the complex process of drug discovery and development? A short anonymous survey was conducted, by invitation, to safety pharmacology representatives of the top 12 pharmaceutical companies, as defined by 2009 revenue figures. A series of multiple choice questions was designed to explore group size, accountabilities, roles and responsibilities of group members, outsourcing policy and publication record. A 92% response rate was obtained. Six out of 11 companies have 10 to 30 full time equivalents in safety pharmacology, who hold similar roles and responsibilities; although the majority of members are not qualified at PhD level or equivalent. Accountabilities were similar across companies and all groups have accountability for core battery in vivo studies and problem solving activities but differences do exist for example with in vitro safety screening and pharmacodynamic/pharmokinetic modeling (PK/PD). The majority of companies outsource less than 25% of studies, with in vitro profiling being the most commonly outsourced activity. Finally, safety pharmacology groups are publishing 1 to 4 articles each year. This short survey has highlighted areas of similarity and differences in the way large pharmaceutical companies operate safety pharmacology. Copyright © 2012 Elsevier Inc. All rights reserved.

Development of an evaluation framework for African-European hospital patient safety partnerships.

PubMed

Rutter, Paul; Syed, Shamsuzzoha B; Storr, Julie; Hightower, Joyce D; Bagheri-Nejad, Sepideh; Kelley, Edward; Pittet, Didier

2014-04-01

Patient safety is recognised as a significant healthcare problem worldwide, and healthcare-associated infections are an important aspect. African Partnerships for Patient Safety is a WHO programme that pairs hospitals in Africa with hospitals in Europe with the objective to work together to improve patient safety. To describe the development of an evaluation framework for hospital-to-hospital partnerships participating in the programme. The framework was structured around the programme's three core objectives: facilitate strong interhospital partnerships, improve in-hospital patient safety and spread best practices nationally. Africa-based clinicians, their European partners and experts in patient safety were closely involved in developing the evaluation framework in an iterative process. The process defined six domains of partnership strength, each with measurable subdomains. We developed a questionnaire to measure these subdomains. Participants selected six indicators of hospital patient safety improvement from a short-list of 22 based on their relevance, sensitivity to intervention and measurement feasibility. Participants proposed 20 measures of spread, which were refined into a two-part conceptual framework, and a data capture tool created. Taking a highly participatory approach that closely involved its end users, we developed an evaluation framework and tools to measure partnership strength, patient safety improvements and the spread of best practice.

Non-invasive, transient determination of the core temperature of a heat-generating solid body

PubMed Central

Anthony, Dean; Sarkar, Daipayan; Jain, Ankur

2016-01-01

While temperature on the surface of a heat-generating solid body can be easily measured using a variety of methods, very few techniques exist for non-invasively measuring the temperature inside the solid body as a function of time. Measurement of internal temperature is very desirable since measurement of just the surface temperature gives no indication of temperature inside the body, and system performance and safety is governed primarily by the highest temperature, encountered usually at the core of the body. This paper presents a technique to non-invasively determine the internal temperature based on the theoretical relationship between the core temperature and surface temperature distribution on the outside of a heat-generating solid body as functions of time. Experiments using infrared thermography of the outside surface of a thermal test cell in a variety of heating and cooling conditions demonstrate good agreement of the predicted core temperature as a function of time with actual core temperature measurement using an embedded thermocouple. This paper demonstrates a capability to thermally probe inside solid bodies in a non-invasive fashion. This directly benefits the accurate performance prediction and control of a variety of engineering systems where the time-varying core temperature plays a key role. PMID:27804981

Non-invasive, transient determination of the core temperature of a heat-generating solid body

NASA Astrophysics Data System (ADS)

Anthony, Dean; Sarkar, Daipayan; Jain, Ankur

2016-11-01

While temperature on the surface of a heat-generating solid body can be easily measured using a variety of methods, very few techniques exist for non-invasively measuring the temperature inside the solid body as a function of time. Measurement of internal temperature is very desirable since measurement of just the surface temperature gives no indication of temperature inside the body, and system performance and safety is governed primarily by the highest temperature, encountered usually at the core of the body. This paper presents a technique to non-invasively determine the internal temperature based on the theoretical relationship between the core temperature and surface temperature distribution on the outside of a heat-generating solid body as functions of time. Experiments using infrared thermography of the outside surface of a thermal test cell in a variety of heating and cooling conditions demonstrate good agreement of the predicted core temperature as a function of time with actual core temperature measurement using an embedded thermocouple. This paper demonstrates a capability to thermally probe inside solid bodies in a non-invasive fashion. This directly benefits the accurate performance prediction and control of a variety of engineering systems where the time-varying core temperature plays a key role.

A study to evaluate the efficacy of image-guided core biopsy in the diagnosis and management of lymphoma--results in 103 biopsies.

PubMed

Vandervelde, C; Kamani, T; Varghese, A; Ramesar, K; Grace, R; Howlett, D C

2008-04-01

The reason for this study was to evaluate the ability of image-guided core biopsy to replace surgical excision by providing sufficient diagnostic and treatment information. All consecutive image-guided core biopsies in patients with a final diagnosis of lymphoma over a 6-year period at our institution were collected retrospectively. Case notes and pathology reports were reviewed and the diagnostic techniques used were recorded. Pathology reports were graded according to their diagnostic completeness and their ability to provide treatment information. Out of a total of 328 instances of lymphoma, 103 image-guided core biopsies were performed in 96 patients. In 78% of these, the diagnostic information obtained from the biopsy provided a fully graded and subtyped diagnosis of lymphoma with sufficient information to initiate therapy. In the head and neck 67% of core biopsies were fully diagnostic for treatment purposes compared to 91% in the thorax, abdomen and pelvis. Image-guided core biopsy has a number of cost and safety advantages over surgical excision biopsy and in suitable cases it can obviate the need for surgery in cases of suspected lymphoma. This is especially relevant for elderly patients and those with poor performance status.

Recovery and Lithologic Analysis of Sediment from Hole UT-GOM2-1-H002, Green Canyon 955, Northern Gulf of Mexico

NASA Astrophysics Data System (ADS)

Kinash, N.; Cook, A.; Sawyer, D.; Heber, R.

2017-12-01

In May 2017 the University of Texas led a drilling and pressure coring expedition in the northern Gulf of Mexico, UT-GOM2-01. The holes were located in Green Canyon Block 955, where the Gulf of Mexico Joint Industry Project Leg II identified an approximately 100m thick hydrate-filled course-grained levee unit in 2009. Two separate wells were drilled into this unit: Holes H002 and H005. In Hole H002, a cutting shoe drill bit was used to collect the pressure cores, and only 1 of the 8 cores collected was pressurized during recovery. The core recovery in Hole H002 was generally poor, about 34%, while the only pressurized core had 45% recovery. In Hole H005, a face bit was used during pressure coring where 13 cores were collected and 9 cores remained pressurized. Core recovery in Hole H005 was much higher, at about 75%. The type of bit was not the only difference between the holes, however. Drilling mud was used throughout the drilling and pressure coring of Hole H002, while only seawater was used during the first 80m of pressure cores collected in Hole H005. Herein we focus on lithologic analysis of Hole H002 with the goal of documenting and understanding core recovery in Hole H002 to compare with Hole H005. X-ray Computed Tomography (XCT) images were collected by Geotek on pressurized cores, mostly from Hole H005, and at Ohio State on unpressurized cores, mostly from Hole H002. The XCT images of unpressurized cores show minimal sedimentary structures and layering, unlike the XCT images acquired on the pressurized, hydrate-bearing cores. Only small sections of the unpressurized cores remained intact. The unpressurized cores appear to have two prominent facies: 1) silt that did not retain original sedimentary fabric and often was loose within the core barrel, and 2) dense mud sections with some sedimentary structures and layering present. On the XCT images, drilling mud appears to be concentrated on the sides of cores, but also appears in layers and fractures within intact core sections. On microscope images, the drilling mud also appears to saturate the pores in some silt intervals. Further analysis of the unpressurized cores is planned, including X-ray diffraction, grain size analysis and porosity measurements. These results will be compared to the pressurized cores to understand if further lithologic factors could have affected core recovery.

Model-Based Safety Analysis

NASA Technical Reports Server (NTRS)

Joshi, Anjali; Heimdahl, Mats P. E.; Miller, Steven P.; Whalen, Mike W.

2006-01-01

System safety analysis techniques are well established and are used extensively during the design of safety-critical systems. Despite this, most of the techniques are highly subjective and dependent on the skill of the practitioner. Since these analyses are usually based on an informal system model, it is unlikely that they will be complete, consistent, and error free. In fact, the lack of precise models of the system architecture and its failure modes often forces the safety analysts to devote much of their effort to gathering architectural details about the system behavior from several sources and embedding this information in the safety artifacts such as the fault trees. This report describes Model-Based Safety Analysis, an approach in which the system and safety engineers share a common system model created using a model-based development process. By extending the system model with a fault model as well as relevant portions of the physical system to be controlled, automated support can be provided for much of the safety analysis. We believe that by using a common model for both system and safety engineering and automating parts of the safety analysis, we can both reduce the cost and improve the quality of the safety analysis. Here we present our vision of model-based safety analysis and discuss the advantages and challenges in making this approach practical.

A model study of the Haihe river passenger ferry risk based on AHP

NASA Astrophysics Data System (ADS)

Du, Jinyin; Xu, Yanming; Du, Chunzhi; Jin, Zhenhua

2017-05-01

The core function of maritime is water safety supervision, whose emphasis and difficulty is ferry. In combination with the practical situation of Haihe river passenger ferry operation management, this paper analyzes Haihe river passenger ferry risk from four aspects "human, machinery, environment and management", and establishes the ferry risk index system. By using AHP (Analytic Hierarchy Process), the ferry risk evaluation model is established. By using the ferry model, the application of Ferry Zhengyanfa7 in Tianjin Haihe river crossing is evaluated, whose safety situation is verified to be between "relatively high risk" and "high risk".

Improvement of Speckle Contrast Image Processing by an Efficient Algorithm.

PubMed

Steimers, A; Farnung, W; Kohl-Bareis, M

2016-01-01

We demonstrate an efficient algorithm for the temporal and spatial based calculation of speckle contrast for the imaging of blood flow by laser speckle contrast analysis (LASCA). It reduces the numerical complexity of necessary calculations, facilitates a multi-core and many-core implementation of the speckle analysis and enables an independence of temporal or spatial resolution and SNR. The new algorithm was evaluated for both spatial and temporal based analysis of speckle patterns with different image sizes and amounts of recruited pixels as sequential, multi-core and many-core code.

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

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

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physicsmore » benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/11 that conserved the mass of each nuclide and volume of each region. The simple cylindrical model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built three-dimensional model and the homogeneous two-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/11 loading 10 to obtain the k{sub eff} for the benchmark model.« less

User’s guide to the collection and analysis of tree cores to assess the distribution of subsurface volatile organic compounds

USGS Publications Warehouse

Vroblesky, Don A.

2008-01-01

Analysis of the volatile organic compound content of tree cores is an inexpensive, rapid, simple approach to examining the distribution of subsurface volatile organic compound contaminants. The method has been shown to detect several volatile petroleum hydrocarbons and chlorinated aliphatic compounds associated with vapor intrusion and ground-water contamination. Tree cores, which are approximately 3 inches long, are obtained by using an increment borer. The cores are placed in vials and sealed. After a period of equilibration, the cores can be analyzed by headspace analysis gas chromatography. Because the roots are exposed to volatile organic compound contamination in the unsaturated zone or shallow ground water, the volatile organic compound concentrations in the tree cores are an indication of the presence of subsurface volatile organic compound contamination. Thus, tree coring can be used to detect and map subsurface volatile organic compound contamination. For comparison of tree-core data at a particular site, it is important to maintain consistent methods for all aspects of tree-core collection, handling, and analysis. Factors affecting the volatile organic compound concentrations in tree cores include the type of volatile organic compound, the tree species, the rooting depth, ground-water chemistry, the depth to the contaminated horizon, concentration differences around the trunk related to variations in the distribution of subsurface volatile organic compounds, concentration differences with depth of coring related to volatilization loss through the bark and possibly other unknown factors, dilution by rain, seasonal influences, sorption, vapor-exchange rates, and within-tree volatile organic compound degradation.

Making the Hubble Space Telescope servicing mission safe

NASA Technical Reports Server (NTRS)

Bahr, N. J.; Depalo, S. V.

1992-01-01

The implementation of the HST system safety program is detailed. Numerous safety analyses are conducted through various phases of design, test, and fabrication, and results are presented to NASA management for discussion during dedicated safety reviews. Attention is given to the system safety assessment and risk analysis methodologies used, i.e., hazard analysis, fault tree analysis, and failure modes and effects analysis, and to how they are coupled with engineering and test analysis for a 'synergistic picture' of the system. Some preliminary safety analysis results, showing the relationship between hazard identification, control or abatement, and finally control verification, are presented as examples of this safety process.

Microbial Performance of Food Safety Control and Assurance Activities in a Fresh Produce Processing Sector Measured Using a Microbial Assessment Scheme and Statistical Modeling.

PubMed

Njage, Patrick Murigu Kamau; Sawe, Chemutai Tonui; Onyango, Cecilia Moraa; Habib, I; Njagi, Edmund Njeru; Aerts, Marc; Molenberghs, Geert

2017-01-01

Current approaches such as inspections, audits, and end product testing cannot detect the distribution and dynamics of microbial contamination. Despite the implementation of current food safety management systems, foodborne outbreaks linked to fresh produce continue to be reported. A microbial assessment scheme and statistical modeling were used to systematically assess the microbial performance of core control and assurance activities in five Kenyan fresh produce processing and export companies. Generalized linear mixed models and correlated random-effects joint models for multivariate clustered data followed by empirical Bayes estimates enabled the analysis of the probability of contamination across critical sampling locations (CSLs) and factories as a random effect. Salmonella spp. and Listeria monocytogenes were not detected in the final products. However, none of the processors attained the maximum safety level for environmental samples. Escherichia coli was detected in five of the six CSLs, including the final product. Among the processing-environment samples, the hand or glove swabs of personnel revealed a higher level of predicted contamination with E. coli , and 80% of the factories were E. coli positive at this CSL. End products showed higher predicted probabilities of having the lowest level of food safety compared with raw materials. The final products were E. coli positive despite the raw materials being E. coli negative for 60% of the processors. There was a higher probability of contamination with coliforms in water at the inlet than in the final rinse water. Four (80%) of the five assessed processors had poor to unacceptable counts of Enterobacteriaceae on processing surfaces. Personnel-, equipment-, and product-related hygiene measures to improve the performance of preventive and intervention measures are recommended.

Characterization Data Package for Containerized Sludge Samples Collected from Engineered Container SCS-CON-210

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

Fountain, Matthew S.; Fiskum, Sandra K.; Baldwin, David L.

This data package contains the K Basin sludge characterization results obtained by Pacific Northwest National Laboratory during processing and analysis of four sludge core samples collected from Engineered Container SCS-CON-210 in 2010 as requested by CH2M Hill Plateau Remediation Company. Sample processing requirements, analytes of interest, detection limits, and quality control sample requirements are defined in the KBC-33786, Rev. 2. The core processing scope included reconstitution of a sludge core sample distributed among four to six 4-L polypropylene bottles into a single container. The reconstituted core sample was then mixed and subsampled to support a variety of characterization activities. Additionalmore » core sludge subsamples were combined to prepare a container composite. The container composite was fractionated by wet sieving through a 2,000 micron mesh and a 500-micron mesh sieve. Each sieve fraction was sampled to support a suite of analyses. The core composite analysis scope included density determination, radioisotope analysis, and metals analysis, including the Waste Isolation Pilot Plant Hazardous Waste Facility Permit metals (with the exception of mercury). The container composite analysis included most of the core composite analysis scope plus particle size distribution, particle density, rheology, and crystalline phase identification. A summary of the received samples, core sample reconstitution and subsampling activities, container composite preparation and subsampling activities, physical properties, and analytical results are presented. Supporting data and documentation are provided in the appendices. There were no cases of sample or data loss and all of the available samples and data are reported as required by the Quality Assurance Project Plan/Sampling and Analysis Plan.« less