Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

NASA Astrophysics Data System (ADS)

Bury, Tomasz

2011-12-01

Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

DOE PAGES

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David; ...

2017-04-30

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

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

Gamble, Kyle A.; Barani, Tommaso; Pizzocri, David

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterionmore » is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Here, further experiments are required to confirm these observations.« less

77 FR 19740 - Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-02

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0249] Water Sources for Long-Term Recirculation Cooling... Regulatory Guide (RG) 1.82, ``Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant... regarding the sumps and suppression pools that provide water sources for emergency core cooling, containment...

Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504. [PWR

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

Booker, C.P.

TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core.

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

NASA Astrophysics Data System (ADS)

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

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

Shum, D.K.M.

This paper examines various issues that would impact the incorporation of warm prestress (WPS) effects in the fracture-margin assessment of reactor pressure vessels (RPVs). By way of an example problem, possible beneficial effects of including type-I WPS in the assessment of an RPV subjected to a small break loss of coolant accident are described. In addition, the need to consider possible loss of constraint effects when interpreting available small specimen WPS-enhanced fracture toughness data is demonstrated through two- and three-dimensional local crack-lip field analyses of a compact tension specimen. Finally, a hybrid correlative-predictive model of WPS base on J-Q theorymore » and the Ritchie-Knott-Rice model is applied to a small scale yielding boundary layer formulation to investigate near crack-tip fields under varying degrees of loading and unloading.« less

A passively-safe fusion reactor blanket with helium coolant and steel structure

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

Crosswait, Kenneth Mitchell

1994-04-01

Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel asmore » a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.« less

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

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

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Warm prestress effects in fracture-margin assessment of PWR-RPVs

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

Shum, D.K.M.

This paper examines various issues that would impact the incorporation of warm prestress (WPS) effects in the fracture-margin assessment of reactor pressure vessels (RPVs). By way of an example problem, possible beneficial effects of including type-I WPS in the assessment of an RPV subjected to a small break loss of coolant accident are described. In addition, the need to consider possible loss of constraint effects when interpreting available small specimen WPS-enhanced fracture toughness data is demonstrated through two- and three-dimensional local crack-lip field analyses of a compact tension specimen. Finally, a hybrid correlative-predictive model of WPS base on J-Q theorymore » and the Ritchie-Knott-Rice model is applied to a small scale yielding boundary layer formulation to investigate near crack-tip fields under varying degrees of loading and unloading.« less

ANALYSIS OF BORON DILUTION TRANSIENTS IN PWRS.

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

DIAMOND,D.J.BROMLEY,B.P.ARONSON,A.L.

2004-02-04

A study has been carried out with PARCS/RELAP5 to understand the consequences of hypothetical boron dilution events in pressurized water reactors. The scenarios of concern start with a small-break loss-of-coolant accident. If the event leads to boiling in the core and then the loss of natural circulation, a boron-free condensate can accumulate in the cold leg. The dilution event happens when natural circulation is re-established or a reactor coolant pump (RCP) is restarted in violation of operating procedures. This event is of particular concern in B&W reactors with a lowered-loop design and is a Generic Safety Issue for the U.S.more » Nuclear Regulatory Commission. The results of calculations with the reestablishment of natural circulation show that there is no unacceptable fuel damage. This is determined by calculating the maximum fuel pellet enthalpy, based on the three-dimensional model, and comparing it with the criterion for damage. The calculation is based on a model of a B&W reactor at beginning of the fuel cycle. If an RCP is restarted, unacceptable fuel damage may be possible in plants with sufficiently large volumes of boron-free condensate in the cold leg.« less

TRAC-PF1/MOD1 support calculations for the MIST/OTIS program

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

Fujita, R.K.; Knight, T.D.

1984-01-01

We are using the Transient Reactor Analysis Code (TRAC), specifically version TRAC-PF1/MOD1, to perform analyses in support of the MultiLoop Integral-System Test (MIST) and the Once-Through Integral-System (OTIS) experiment program. We have analyzed Geradrohr Dampferzeuger Anlage (GERDA) Test 1605AA to benchmark the TRAC-PF1/MOD1 code against phenomena expected to occur in a raised-loop B and W plant during a small-break loss-of-coolant accident (SBLOCA). These results show that the code can calculate both single- and two-phase natural circulation, flow interruption, boiler-condenser-mode (BCM) heat transfer, and primary-system refill in a B and W-type geometry with low-elevation auxiliary feedwater. 19 figures, 7 tables.

Simulation of a small cold-leg-break experiment at the PMK-2 test facility using the RELAP5 and ATHLET codes

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

Ezsoel, G.; Guba, A.; Perneczky, L.

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of themore » experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.« less

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

Gerard, R.; Malekian, C.; Meessen, O.

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports inmore » the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.« less

Material distribution in light water reactor-type bundles tested under severe accident conditions

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

Noack, V.; Hagen, S.J.L.; Hofmann, P.

1997-02-01

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorbermore » and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.« less

Experimental study of phase separation in dividing two phase flow

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

Qian Yong; Yang Zhilin; Xu Jijun

1996-12-31

Experimental study of phase separation of air-water two phase bubbly, slug flow in the horizontal T-junction is carried out. The influences of the inlet mass quality X1, mass extraction rate G3/G1, and fraction of extracted liquid QL3/QL1 on phase separation characteristics are analyzed. For the first time, the authors have found and defined pulsating run effect by the visual experiments, which show that under certain conditions, the down stream flow of the T-junction has strangely affected the phase redistribution of the junction, and firstly point out that the downstream geometric condition is very important to the study of phase separationmore » phenomenon of two-phase flow in a T-junction. This kind of phenomenon has many applications in the field of energy, power, petroleum and chemical industries, such as the loss of coolant accident (LOCA) caused by a small break in a horizontal coolant pipe in nuclear reactor, and the flip-flop effect in the natural gas transportation pipeline system, etc.« less

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

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

Hamm, L.L.

1998-10-07

This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

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

49 CFR 178.345-8 - Accident damage protection.

Code of Federal Regulations, 2010 CFR

2010-10-01

... accidental loss of lading. The device must break at no more than 70 percent of the load that would be... major radius of the tank shell. The device must break at no more than 70 percent of the load that would... for the loss of lading due to an accident. (1) Any dome, sump, or washout cover plate projecting from...

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

Decay Heat Removal from a GFR Core by Natural Convection

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

Williams, Wesley C.; Hejzlar, Pavel; Driscoll, Michael J.

2004-07-01

One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work investigates the potential of post-LOCA decay heat removal from a GFR core to a heat sink using an external convection loop. A model was developed in the form of the LOCA-COLA (Loss of Coolant Accident - Convection Loop Analysis) computer code as a meansmore » for 1D steady state convective heat transfer loop analysis. The results show that decay heat removal by means of gas cooled natural circulation is feasible under elevated post-LOCA containment pressure conditions. (authors)« less

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

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

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

77 FR 65720 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-30

... transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming... hot channel factor, nuclear enthalpy rise hot channel factor, loss of coolant accident peak cladding...

Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

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

Kervinen, T.; Riikonen, V.; Ritonummi, T.

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing themore » heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.« less

76 FR 66333 - Carolina Power & Light Company, H.B. Robinson Steam Electric Plant, Unit No. 2; Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-26

..., and hydrogen generation after a postulated loss-of-coolant accident. Therefore, both of these... quality. There are no impacts to historical and cultural resources. In addition, there are also no known...

Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

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

Vigil, R.A.; Jacobus, M.J.

1994-04-01

Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables,more » the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent.« less

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

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Solar concentrator protective system

NASA Technical Reports Server (NTRS)

Selcuk, M. K. (Inventor)

1984-01-01

A mechanism that blocks concentrated sunlight from reaching a receiver, in the event of a tracking failure or loss of coolant is described. Sunlight is normally concentrated by a dish reflector onto the opening of a receiver. A faceplate surrounds the opening, and coolant carrying tubes, line the receiver. If the concentrated sunlight wanders so it begins to fall on the faceplate, then the sunlight will melt a portion of a fuse wire portion will break. The wire is attached to a flange on a shutter frame, and breaking of the fuse wire allows the frame to fall. Normally, the shutter frame supports shutter elements that are held open by cam followers that bear against cams.

Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

NASA Astrophysics Data System (ADS)

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

2016-12-01

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

Loss of control air at Browns Ferry Unit One: accident sequence analysis

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

Harrington, R.M.; Hodge, S.A.

1986-04-01

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure injection systems. Accident sequence calculations are presented for Loss of Control Air sequences with assumed failure upon demand of the Reactor Core Isolation Cooling (RCIC) and the High Pressure Coolant Injection (HPCI) at Unit 1. Sequences with and without operator action are considered.more » Results show that the operators can prevent core uncovery if they take action to utilize the Control Rod Drive Hydraulic System as a backup high pressure injection system.« less

Time-to-burnout data for a prototypical ITER divertor tube during a simulated loss of flow accident

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

Marshall, T.D.; Watson, R.D.; McDonald, J.M.

The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm.more » The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. After reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube`s heated surface exhibited the high temperature transient normally indicative of rapidly approaching burnout. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux.« less

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

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

Steiner, J.L.; Harmony, S.C.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS Supplement design were performed for a large-break loss-of-coolant (LBLOCA) initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following anmore » LBLOCA.« less

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Development and Validation of Accident Models for FeCrAl Cladding

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to present the work completed in regards to material model development for FeCrAl cladding and highlight the results of applying these models to Loss of Coolant Accidents (LOCA) and Station Blackouts (SBO). With the limited experimental data available (essentially only the data used to create the models) true validation is not possible. In the absence of another alternative, qualitative comparisons during postulated accident scenarios between FeCrAl and Zircaloy-4 cladded rods have been completed demonstrating the superior performance of FeCrAl.

Thermal Stratification Analysis for Sodium Fast Reactors

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

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

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

Baek, J. S.; Cheng, L. Y.; Diamond, D.

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enrichedmore » uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.« less

CFD Analyses of Air-Ingress Accident for VHTRs

NASA Astrophysics Data System (ADS)

Ham, Tae Kyu

The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air-ingress mechanism and to utilize the CFD simulation in the analysis of the phenomenon. Previous air-ingress studies simulated the depressurization process using simple assumptions or 1-D system code results. However, recent studies found flow oscillations near the end of the depressurization which could influence the next stage of the air-ingress accident. Therefore, CFD simulations were performed to examine the air-ingress mechanisms from the depressurization through the establishment of local natural circulation initiate. In addition to the double-guillotine break scenario, there are other scenarios that can lead to an air-ingress event such as a partial break were in the cross vessel with various break locations, orientations, and shapes. These additional situations were also investigated. The simulation results for the OSU test facility showed that the discharged helium coolant from a reactor vessel during the depressurization process will be mixed with the air in the containment. This process makes the density of the gas mixture in the containment lower and the density-driven air-ingress flow slower because the density-driven flow is established by the density difference of the gas species between the reactor vessel and the containment. In addition, for the simulations with various initial and boundary conditions, the simulation results showed that the total accumulated air in the containment collapsed within 10% standard deviation by: 1. multiplying the density ratio and viscosity ratio of the gas species between the containment and the reactor vessel and 2. multiplying the ratio of the air mole fraction and gas temperature to the reference value. By replacing the gas mixture in the reactor cavity with a gas heavier than the air, the air-ingress speed slowed down. Based on the understanding of the air-ingress phenomena for the GT-MHR air-ingress scenario, several mitigation measures of air-ingress accident are proposed. The CFD results are utilized to plan experimental strategy and apparatus installation to obtain the best results when conducting an experiment. The validation of the generated CFD solutions will be performed with the OSU air-ingress experimental results. (Abstract shortened by UMI.).

Summary on the depressurization from supercritical pressure conditions

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

Anderson, M.; Chen, Y.; Ammirable, L.

When a fluid discharges from a high pressure and temperature system, a 'choking' or critical condition occurs, and the flow rate becomes independent of the downstream pressure. During a postulated loss of coolant accident (LOCA) of a water reactor the break flow will be subject to this condition. An accurate estimation of the critical flow rate is important for the evaluation of the reactor safety, because this flow rate controls the loss of coolant inventory and energy from the system, and thus has a significant effect on the accident consequences[1]. In the design of safety systems for a super criticalmore » water reactor (SCWR), postulated LOCA transients are particularly important due to the lower coolant inventory compared to a typical PWR for the same power output. This lower coolant inventory would result in a faster transient response of the SCWR, and hence accurate prediction of the critical discharge is mandatory. Under potential two-phase conditions critical flow is dominated by the vapor content or quality of the vapor, which is closely related with the onset of vaporization and the interfacial interaction between phases [2]. This presents a major challenge for the estimation of the flow rate due to the lack of the knowledge of those processes, especially under the conditions of interest for the SCWR. According to the limited data of supercritical fluids, the critical flows at conditions above the pseudo-critical point seem to be fairly stable and consistent with the subcritical homogeneous equilibrium model (HEM) model predictions, while having a lower flow rate than those in the two-phase region. Thus the major difficulty in the prediction of the depressurization flow rates remains in the region where two phases co-exist at the top of the vapor dome. In this region, the flow rate is strongly affected by the nozzle geometry and tends to be unstable. Various models for this region have been developed with different assumptions, e.g. the HEM and Moody model [3], and the Henry-Fauske non-equilibrium model [4], and are currently used in subcritical pressure reactor safety design[5]. It appears that some of these models could be reasonably extended to above the thermodynamic pseudo-critical point. The more stable and lower discharge flow rates observed in conditions above the pseudo-critical point suggests that even though SCWR's have a smaller coolant inventory, the safety implications of a LOCA and the subsequent depressurization may not be as severe as expected, this however needs to be confirmed by a rigorous evaluation of the particular event and further evaluation of the critical flow rate. This paper will summarize activities on critical flow models, experimental data and numerical modeling during blowdown from supercritical pressure conditions under the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on 'Heat Transfer Behaviour and Thermo-hydraulics Code testing for SCWRs'. (authors)« less

Emergency cooling analysis for the loss of coolant malfunction

NASA Technical Reports Server (NTRS)

Peoples, J. A.

1972-01-01

This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.

Analysis of gamma ray dose for dried up pond storing low enriched UO2 fuel

NASA Astrophysics Data System (ADS)

Nauchi, Yasushi; Suzuki, Motomu

2017-09-01

Gamma ray dose is calculated for loss of coolant accident in spent fuel pond (SFP) storing irradiated fuels used in light water reactors. Influence of modelling of fuel assemblies, source distributions, and loading fraction of fuel assemblies in the fuel rack on the dose are investigated.

77 FR 53923 - Biweekly Notice;

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-04

... to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well... (psig) to 49.7 psig for the design basis loss-of- coolant accident (LOCA). In support of the revised P a... analysis. The P a remains below the containment design pressure of 50 psig because of the change in the...

PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

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

Sienicki, J.J.

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executedmore » on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.« less

49 CFR 178.345-8 - Accident damage protection.

Code of Federal Regulations, 2011 CFR

2011-10-01

... accidental loss of lading. The device must break at no more than 70 percent of the load that would be... major radius of the tank shell. The device must break at no more than 70 percent of the load that would... requirements of this section and the applicable individual specification to minimize the potential for the loss...

49 CFR 178.345-8 - Accident damage protection.

Code of Federal Regulations, 2013 CFR

2013-10-01

... accidental loss of lading. The device must break at no more than 70 percent of the load that would be... major radius of the tank shell. The device must break at no more than 70 percent of the load that would... requirements of this section and the applicable individual specification to minimize the potential for the loss...

49 CFR 178.345-8 - Accident damage protection.

Code of Federal Regulations, 2014 CFR

2014-10-01

... accidental loss of lading. The device must break at no more than 70 percent of the load that would be... major radius of the tank shell. The device must break at no more than 70 percent of the load that would... requirements of this section and the applicable individual specification to minimize the potential for the loss...

49 CFR 178.345-8 - Accident damage protection.

Code of Federal Regulations, 2012 CFR

2012-10-01

... accidental loss of lading. The device must break at no more than 70 percent of the load that would be... major radius of the tank shell. The device must break at no more than 70 percent of the load that would... requirements of this section and the applicable individual specification to minimize the potential for the loss...

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

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

Franken, Daniel; Gould, Daniel; Jain, Prashant K.

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE PAGES

Franken, Daniel; Gould, Daniel; Jain, Prashant K.; ...

2017-09-21

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Extravehicular Mobility Unit (EMU) / International Space Station (ISS) Coolant Loop Failure and Recovery

NASA Technical Reports Server (NTRS)

Lewis, John F.; Cole, Harold; Cronin, Gary; Gazda, Daniel B.; Steele, John

2006-01-01

Following the Colombia accident, the Extravehicular Mobility Units (EMU) onboard ISS were unused for several months. Upon startup, the units experienced a failure in the coolant system. This failure resulted in the loss of Extravehicular Activity (EVA) capability from the US segment of ISS. With limited on-orbit evidence, a team of chemists, engineers, metallurgists, and microbiologists were able to identify the cause of the failure and develop recovery hardware and procedures. As a result of this work, the ISS crew regained the capability to perform EVAs from the US segment of the ISS.

75 FR 36698 - Draft Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... information based on the likelihood of pipe breaks of different sizes. The rule would divide all coolant... to and including a ``transition break size,'' and breaks larger than the transition size up to the largest pipe in the reactor coolant system. Selection of the transition size was based upon pipe break...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

Light-water-reactor safety research program. Quarterly progress report, July--September 1975

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

Not Available

1975-01-01

Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

Fast reactor safety and related physics. Volume IV. Phenomenology

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

Not Available

1976-01-01

Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

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

Posttest analysis of MIST Test 320201 using TRAC-PF1/MOD1

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

Siebe, D.A.; Steiner, J.L.; Boyack, B.E.

A posttest calculation and analysis of Multi-Loop Integral System Test 320201, a small-break loss-of-coolant accident (SBLOCA) test with a scaled 50-cm{sup 2} cold-leg pump discharge leak, has been completed and is reported herein. It was one in a series of tests, with leak size varied parametrically. Scaled leak sizes included 5, 10, (the nominal, Test 3109AA), and 50 cm{sub 2}. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, interruption of loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vesselmore » vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator (SG) secondaries was used after SG-secondary refill; and symmetric SG pressure control was also used. The sequence of events seen in this test was similar to the sequence of events for much of the nominal test except that events occurred in a shorter time frame as the system inventory was reduced and the system depressurized at a faster rate. The calculation was performed using TRAC-PFL/MOD 1. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Posttest analysis of MIST Test 320201 using TRAC-PF1/MOD1

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

Siebe, D.A.; Steiner, J.L.; Boyack, B.E.

A posttest calculation and analysis of Multi-Loop Integral System Test 320201, a small-break loss-of-coolant accident (SBLOCA) test with a scaled 50-cm[sup 2] cold-leg pump discharge leak, has been completed and is reported herein. It was one in a series of tests, with leak size varied parametrically. Scaled leak sizes included 5, 10, (the nominal, Test 3109AA), and 50 cm[sub 2]. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, interruption of loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vesselmore » vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator (SG) secondaries was used after SG-secondary refill; and symmetric SG pressure control was also used. The sequence of events seen in this test was similar to the sequence of events for much of the nominal test except that events occurred in a shorter time frame as the system inventory was reduced and the system depressurized at a faster rate. The calculation was performed using TRAC-PFL/MOD 1. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

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

Faidy, C.; Gilles, P.

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of:more » simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.« less

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

Multiloop Integral System Test (MIST): MIST Facility Functional Specification

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

Habib, T F; Koksal, C G; Moskal, T E

1991-04-01

The Multiloop Integral System Test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility --more » the Once Through Integral System (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST Functional Specification documents as-built design features, dimensions, instrumentation, and test approach. It also presents the scaling basis for the facility and serves to define the scope of work for the facility design and construction. 13 refs., 112 figs., 38 tabs.« less

Physics of some environmental aspects of energy

NASA Astrophysics Data System (ADS)

Hafemeister, David

1985-11-01

Approximate numerical estimates are carried out on the following environmental effects from energy production and conservation: (1) The greenhouse effect caused by increased CO2 in the atmosphere; (2) Loss of coolant accidents in nuclear reactors; (3) Increased radon concentrations in buildings with very low air infiltration rates; (4) Acid rain from the combustion of fossil fuels; and (5) Explosions of liquified natural gas (LNG).

Hot Cell Installation and Demonstration of the Severe Accident Test Station

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

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« less

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

Pickman, D.O.

Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

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

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

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

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approachmore » to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)« less

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

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

Mendler, O J; Takeuchi, K; Young, M Y

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

An Experimental Examination of the Loss-of-Flow Accident Phenomenon for Prototypical ITER Divertor Channels of Y = 0 and Y = 2

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

Marshall, Theron D.; McDonald, Jimmie M.; Cadwallader, Lee C.

2000-01-15

This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pumpmore » was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.« less

An experimental examination of the loss-of-flow accident phenomenon for prototypical ITER divertor channels of Y=0 and Y=2

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

Marshall, T.D.; McDonald, J.M.; Cadwallader, L.C.

2000-01-01

This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pumpmore » was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.« less

Condensation model for the ESBWR passive condensers

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

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

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

Mueftueoglu, A.K.; Feltus, M.A.

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and othermore » signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.« less

Reactor vessel lower head integrity

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

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less

Development of a coolant channel helium and nitrogen gas ratio sensor for a high temperature gas reactor

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

Cadell, S. R.; Woods, B. G.

2012-07-01

To measure the changing gas composition of the coolant during a postulated High Temperature Gas Reactor (HTGR) accident, an instrument is needed. This instrument must be compact enough to measure the ratio of the coolant versus the break gas in an individual coolant channel. This instrument must minimally impact the fluid flow and provide for non-direct signal routing to allow minimal disturbance to adjacent channels. The instrument must have a flexible geometry to allow for the measurement of larger volumes such as in the upper or lower plenum of a HTGR. The instrument must be capable of accurately functioning throughmore » the full operating temperature and pressure of a HTGR. This instrument is not commercially available, but a literature survey has shown that building off of the present work on Capacitance Sensors and Cross-Capacitors will provide a basis for the development of the desired instrument. One difficulty in developing and instrument to operate at HTGR temperatures is acquiring an electrical conductor that will not melt at 1600 deg. C. This requirement limits the material selection to high temperature ceramics, graphite, and exotic metals. An additional concern for the instrument is properly accounting for the thermal expansion of both the sensing components and the gas being measured. This work covers the basic instrument overview with a thorough discussion of the associated uncertainty in making these measurements. (authors)« less

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

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

Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin

Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focusedmore » on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.« less

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

Walston, S; Rowland, M; Campbell, K

It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting inmore » a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.« less

Indirect-cycle FBR cooled by supercritical steam-concept and design

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

Yoshiaki, Oka; Tatjana, Jevremovic; Sei-ichi, Koshizuka

1993-01-01

Neutronic and thermal-hydraulic design of an in direct-cycle supercritical steam-cooled fast breeder reactor (SCFBR-I) is carried out to find a way to make low-cost FBRs (Ref. 1). The advantages of supercritical steam cooling are high thermal efficiency, low pumping power, simplified system (no primary steam generators and no Loeffler boilers), and the use of experienced technology in fossil-fired power plants. The design goals are fissile fuel breeding (compound system doubling time below 30 yr), 1000-M(electric) class out-put, high fuel discharge burnup, and a long refueling period. The coolant void reactivity should be negative throughout fuel lifetime because the loss-of-coolant accidentmore » is the design-basis accident. These goals have never been satisfied simultaneously in previous SCFBRs.« less

Core cooling under accident conditions at the high-flux beam reactor

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

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

NASA Astrophysics Data System (ADS)

Ha, Taesung

A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential usefulness of quantifying model uncertainty as sensitivity analysis in the PRA model.

Analysis of helium purification system capability during water ingress accident in RDE

NASA Astrophysics Data System (ADS)

Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Westinghouse Small Modular Reactor passive safety system response to postulated events

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

Smith, M. C.; Wright, R. F.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. Themore » integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)« less

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

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

Feldman, E. M.; Patton, Jr., M. L.; Sackett, K. E.

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulatemore » emergency core coolant injection in a PWR, with the flow rate based on system volume scaling.« less

A defense in depth approach for nuclear power plant accident management

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

Chih-Yao Hsieh; Hwai-Pwu Chou

2015-07-01

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identifymore » what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident management. (authors)« less

TRAC-PF1/MOD1 pretest predictions of MIST experiments

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

Boyack, B.E.; Steiner, J.L.; Siebe, D.A.

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 to provide integral system test data on specific issues and phenomena relevant to post small-break loss-of-coolant accidents (SBLOCAs) in Babcock and Wilcox plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. During Fiscal Year 1986, Los Alamos performed five MIST pretest analyses. The five experiments were chosen on the basis of their potential either to approach the facility limits or to challenge the predictive capability of the TRAC-PF1/MOD1 code. Three SBLOCA tests weremore » examined which included nominal test conditions, throttled auxiliary feedwater and asymmetric steam-generator cooldown, and reduced high-pressure-injection (HPI) capacity, respectively. Also analyzed were two ''feed-and-bleed'' cooling tests with reduced HPI and delayed HPI initiation. Results of the tests showed that the MIST facility limits would not be approached in the five tests considered. Early comparisons with preliminary test data indicate that the TRAC-PF1/MOD1 code is correctly calculating the dominant phenomena occurring in the MIST facility during the tests. Posttest analyses are planned to provide a quantitative assessment of the code's ability to predict MIST transients.« less

Core cooling under accident conditions at the high flux beam reactor (HFBR)

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

Tichler, P.; Cheng, L.; Fauske, H.

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. In a LOCA scenario where even limited fuelmore » damage occurs and natural circulation is established, fission product gases could be carried from the damaged fuel by steam into areas where operator access is required to maintain the core in a coolable configuration. This would force evacuation of the building and lead to extensive core damage. As a result the HFBR was shut down by the Department of Energy (DOE) and an extensive review of the HFBR was initiated. In an effort to address this issue BNL developed a model designed to predict the heat removal limit during flow reversal that was found to be in good agreement with the test results. Currently a thermal-hydraulic test program is being developed to provide a more realistic and defensible estimate of the flow reversal heat removal limit so that the reactor power level can be increased.« less

Safety research of insulating materials of cable for nuclear power generating station

NASA Technical Reports Server (NTRS)

Lee, C. K.; Choi, J. H.; Kong, Y. K.; Chang, H. S.

1988-01-01

The polymers PE, EPR, PVC, Neoprene, CSP, CLPE, EP and other similar substances are frequently used as insulation and protective covering for cables used in nuclear power generating stations. In order to test these materials for flame retardation, environmental resistance, and cable specifications, they were given the cable normal test, flame test, chemical tests, and subjected to design analysis and loss of coolant accident tests. Material was collected on spark tests and actual experience standards were established through these contributions and technology was accumulated.

Posttest REALP4 analysis of LOFT experiment L1-3A

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

White, J.R.; Holmstrom, H.L.O.

This report presents selected results of posttest RELAP4 modeling of LOFT loss-of-coolant experiment L1-3A, a double-ended isothermal cold leg break with lower plenum emergency core coolant injection. Comparisons are presented between the pretest prediction, the posttest analysis, and the experimental data. It is concluded that pressurizer modeling is important for accurately predicting system behavior during the initial portion of saturated blowdown. Using measured initial conditions rather than nominal specified initial conditions did not influence the system model results significantly. Using finer nodalization in the reactor vessel improved the prediction of the system pressure history by minimizing steam condensation effects. Unequalmore » steam condensation between the downcomer and core volumes appear to cause the manometer oscillations observed in both the pretest and posttest RELAP4 analysis.« less

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

NASA Astrophysics Data System (ADS)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

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

Hsu, M.T.; Davis, C.B.; Behling, S.R.

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio wasmore » maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).« less

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

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

Baratta, A.J.

1997-07-01

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

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

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

Tusheva, P.; Schaefer, F.; Kliem, S.

2012-07-01

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safetymore » systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)« less

Science and society test VII: Energy and environment

NASA Astrophysics Data System (ADS)

Hafemeister, David W.

1982-08-01

Approximate numerical estimates are developed in order to quantify a variety of environmental effects that result from energy production. The results of these calculations are consistent with either direct observations or with more complex calculations. This paper will cover some of the possible environmental effects of the following: (1) the greenhouse effect caused by increased CO2 in the atmosphere; (2) loss of coolant accidents in nuclear reactors; (3) increased radon concentrations in buildings with very low air infiltration rates; (4) acid rain from the combustion of fossil fuels; (5) expolosions of liquified natural gas (LNG); and (6) ozone in the stratosphere.

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

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

Experimental and code simulation of a station blackout scenario for APR1400 with test facility ATLAS and MARS code

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

Yu, X. G.; Kim, Y. S.; Choi, K. Y.

2012-07-01

A SBO (station blackout) experiment named SBO-01 was performed at full-pressure IET (Integral Effect Test) facility ATLAS (Advanced Test Loop for Accident Simulation) which is scaled down from the APR1400 (Advanced Power Reactor 1400 MWe). In this study, the transient of SBO-01 is discussed and is subdivided into three phases: the SG fluid loss phase, the RCS fluid loss phase, and the core coolant depletion and core heatup phase. In addition, the typical phenomena in SBO-01 test - SG dryout, natural circulation, core coolant boiling, the PRZ full, core heat-up - are identified. Furthermore, the SBO-01 test is reproduced bymore » the MARS code calculation with the ATLAS model which represents the ATLAS test facility. The experimental and calculated transients are then compared and discussed. The comparison reveals there was malfunction of equipments: the SG leakage through SG MSSV and the measurement error of loop flow meter. As the ATLAS model is validated against the experimental results, it can be further employed to investigate the other possible SBO scenarios and to study the scaling distortions in the ATLAS. (authors)« less

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

MIST final report

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

Gloudemans, J.R.

1991-08-01

The multiloop integral system test (MIST) was part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock Wilcox-designed plants. MIST was sponsored by the US Nuclear Regulatory Commission, the Babcock Wilcox Owners Group, the Electric Power Research Institute, and Babcock Wilcox. The unique features of the Babcock Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral system facilities to addresss the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility -- the once-through integralmore » system (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST program is reported in eleven volumes; Volumes 2 through 8 pertain to groups of Phase 3 tests by type, Volume 9 presents inter-group comparisons. Volume 10 provides comparisons between the RELAP5 MOD2 calculations and MIST observations, and Volume 11 (with addendum) presents the later, Phase 4 tests. This is Volume 1 of the MIST final report, a summary of the entire MIST program. Major topics include: test advisory grop (TAG) issues; facility scaling and design; test matrix; observations; comparisons of RELAP5 calculations to MIST observations; and MIST versus the TAG issues. 11 refs., 29 figs., 9 tabs.« less

Effects of the air–steam mixture on the permeability of damaged concrete

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

Medjigbodo, Sonagnon; Darquennes, Aveline; Aubernon, Corentin

Massive concrete structures such as the containments of nuclear power plant must maintain their tightness at any circumstances to prevent the escape of radioactive fission products into the environment. In the event of an accident like a Loss of Coolant Accident (LOCA), the concrete wall is submitted to both hydric and mechanical loadings. A new experimental device reproducing these extreme conditions (water vapor transfer, 140 °C and 5 bars) is developed in the GeM Laboratory to determine the effect of the saturation degree, the mechanical loading and the flowing fluid type on the concrete transfer properties. The experimental tests showmore » that the previous parameters significantly affect the concrete permeability and the gas leakage rate. Their evolution as a function of the mechanical loading is characterized by two phases that are directly related to concrete microstructure and crack development.« less

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

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

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

Methods for the mitigation of the chemical reactivity of beryllium in steam

NASA Astrophysics Data System (ADS)

Druyts, F.; Alves, E. C.; Wu, C. H.

2004-08-01

In the safety assessment of future fusion reactors, the reaction of beryllium with steam remains one of the main concerns. In case of a loss of coolant accident (LOCA), the use of beryllium in combination with pressurised water as coolant can lead to excessive hydrogen production due to the reaction Be + H 2O = BeO + H 2 + heat. Therefore, we started an R&D programme aimed at investigating mitigation methods for the beryllium/steam reaction. Beryllium samples were implanted with either calcium or aluminium ions in a 210 kV ion implanter at ITN Lisbon. The chemical reactivity of these samples in steam was measured at SCK • CEN in a dedicated experimental facility providing coupled thermogravimetry/mass spectrometry. In comparison to reference undoped material, the reactivity of doped beryllium after 30 min of exposure decreased with a factor 2 to 4. The mitigating effect was higher for calcium-doped than for aluminium-doped samples.

ATWS at Browns Ferry Unit One - accident sequence analysis

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

Harrington, R.M.; Hodge, S.A.

1984-07-01

This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence themore » quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Effect of steam generator configuration in a loss of the RHR during mid-loop operation at PKL facility

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

Villanueva, J. F.; Carlos, S.; Martorell, S.

The loss of the residual heat removal system in mid-loop conditions may occur with a non-negligible contribution to the plant risk, so the analysis of the accidental sequences and the actions to mitigate the accident are of great interest in shutdown conditions. In order to plan the appropriate measures to mitigate the accident is necessary to understand the thermal-hydraulic processes following the loss of the residual heat removal system during shutdown. Thus, transients of this kind have been simulated using best-estimate codes in different integral test facilities and compared with experimental data obtained in different facilities. In PKL (Primaerkreislauf-Versuchsanlage, primarymore » coolant loop test facility) test facility different series of experiments have been undertaken to analyze the plant response in shutdown. In this context, the E3 and F2 series consist of analyzing the loss of the residual heat removal system with a reduced inventory in the primary system. In particular, the experiments were developed to investigate the influence of the steam generators secondary side configuration on the plant response, what involves the consideration of different number of steam generators filled with water and ready for activation, on the heat transfer mechanisms inside the steam generators U-tubes. This work presents the results of such experiments calculated using, RELAP5/Mod 3.3. (authors)« less

Float level switch for a nuclear power plant containment vessel

DOEpatents

Powell, J.G.

1993-11-16

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

Float level switch for a nuclear power plant containment vessel

DOEpatents

Powell, James G.

1993-01-01

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

Interfacial heat transfer in multiphase molten pools with gas injection

NASA Astrophysics Data System (ADS)

Bilbao Y Leon, Rosa Marina

1998-12-01

In the very unlikely event of a severe reactor accident involving core meltdown and pressure vessel failure, it is vital to identify the circumstances that would allow the molten core material to cool down and resolidify, bringing core debris to a safe and stable state. In this type of accident, the molten material which escapes from the reactor pressure vessel will accumulate as a molten pool in the reactor cavity below. To achieve coolability of the corium in this configuration it has been proposed to flood the cavity with water from above forming a layered structure where upward heat loss from the molten pool to the water will cause the core material to quench and solidify. The effectiveness of this procedure depends largely on the rate of upward heat loss as well as on the formation and stability of an upper crust. In this situation the molten pool becomes a three phase mixture: the solid and liquid slurry formed by the molten pool cooled to a temperature below the temperature of liquidus, agitated by the gases formed in the concrete ablation process. The present work quantifies the partition of the heat losses upward and downward considering the influence of the solid fraction in the pool and the viscosity effects, and the rate of heat loss through a solid layer. To complete this task a intermediate scale experimental test section has been designed and built at the University of Wisconsin - Madison, in which simulant materials are used to model the process of heat and mass transfer which involves the molten pool, the solid layer atop and the coolant layer above. The design includes volumetric heating, gas injection from the bottom and solids within the pool. New experimental results showing the heat transfer behavior for pools with different viscosities and various solid fractions are presented. The current results indicate a power split which favors heat transfer upward to the coolant simulant above by a 2:1 or 3:1 ratio. In addition, the power split is unaffected by the viscosity of the pool, the solid fractions in the pool and the superficial velocity.

Inexpensive system protects megawatt resistance-heating furnace against high-voltage surges

NASA Technical Reports Server (NTRS)

Stearns, E. J.

1971-01-01

Coolant gas extinguishes arcing across the break in a heater element. Air-gap shunt which bypasses high voltage impressed across the circuit prevents damage if the resistance elements break and open the inductive circuit.

Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1

PubMed Central

Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

2015-01-01

In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by 137Cesium (137Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as 132Te-132I, 131I, 134Cs and 137Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h−1 per initial 137Cs deposition of 1000 kBq m−2, whereas it was 100 μGy h−1 around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m−2 for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums (134Cs + 137Cs) around Chernobyl and Fukushima-1, respectively. PMID:26568603

Advanced smart tungsten alloys for a future fusion power plant

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Mao, Y.; Coenen, J. W.; Bram, M.; Gonzalez-Julian, J.

2017-06-01

The severe particle, radiation and neutron environment in a future fusion power plant requires the development of advanced plasma-facing materials. At the same time, the highest level of safety needs to be ensured. The so-called loss-of-coolant accident combined with air ingress in the vacuum vessel represents a severe safety challenge. In the absence of a coolant the temperature of the tungsten first wall may reach 1200 °C. At such a temperature, the neutron-activated radioactive tungsten forms volatile oxide which can be mobilized into atmosphere. Smart tungsten alloys are being developed to address this safety issue. Smart alloys should combine an acceptable plasma performance with the suppressed oxidation during an accident. New thin film tungsten-chromium-yttrium smart alloys feature an impressive 105 fold suppression of oxidation compared to that of pure tungsten at temperatures of up to 1000 °C. Oxidation behavior at temperatures up to 1200 °C, and reactivity of alloys in humid atmosphere along with a manufacturing of reactor-relevant bulk samples, impose an additional challenge in smart alloy development. First exposures of smart alloys in steady-state deuterium plasma were made. Smart tungsten-chroimium-titanium alloys demonstrated a sputtering resistance which is similar to that of pure tungsten. Expected preferential sputtering of alloying elements by plasma ions was confirmed experimentally. The subsequent isothermal oxidation of exposed samples did not reveal any influence of plasma exposure on the passivation of alloys.

Once-through integral system (OTIS): Final report

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

Gloudemans, J R

1986-09-01

A scaled experimental facility, designated the once-through integral system (OTIS), was used to acquire post-small break loss-of-coolant accident (SBLOCA) data for benchmarking system codes. OTIS was also used to investigate the application of the Abnormal Transient Operating Guidelines (ATOG) used in the Babcock and Wilcox (B and W) designed nuclear steam supply system (NSSS) during the course of an SBLOCA. OTIS was a single-loop facility with a plant to model power scale factor of 1686. OTIS maintained the key elevations, approximate component volumes, and loop flow resistances, and simulated the major component phenomena of a B and W raised-loop nuclearmore » plant. A test matrix consisting of 15 tests divided into four categories was performed. The largest group contained 10 tests and was defined to parametrically obtain an extensive set of plant-typical experimental data for code benchmarking. Parameters such as leak size, leak location, and high-pressure injection (HPI) shut-off head were individually varied. The remaining categories were specified to study the impact of the ATOGs (2 tests), to note the effect of guard heater operation on observed phenomena (2 tests), and to provide a data set for comparison with previous test experience (1 test). A summary of the test results and a detailed discussion of Test 220100 is presented. Test 220100 was the nominal or reference test for the parametric studies. This test was performed with a scaled 10-cm/sup 2/ leak located in the cold leg suction piping.« less

Thermal-hydraulic analysis capabilities and methods development at NYPA

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

Feltus, M.A.

1987-01-01

The operation of a nuclear power plant must be regularly supported by various thermal-hydraulic (T/H) analyses that may include final safety analysis report (FSAR) design basis calculations and licensing evaluations and conservative and best-estimate analyses. The development of in-house T/H capabilities provides the following advantages: (a) it leads to a better understanding of the plant design basis and operating characteristics; (b) methods developed can be used to optimize plant operations and enhance plant safety; (c) such a capability can be used for design reviews, checking vendor calculations, and evaluating proposed plant modifications; and (d) in-house capability reduces the cost ofmore » analysis. This paper gives an overview of the T/H capabilities and current methods development activity within the engineering department of the New York Power Authority (NYPA) and will focus specifically on reactor coolant system (RCS) transients and plant dynamic response for non-loss-of-coolant accident events. This paper describes NYPA experience in performing T/H analyses in support of pressurized water reactor plant operation.« less

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

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

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

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

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface

NASA Astrophysics Data System (ADS)

Wu, A.; Ribis, J.; Brachet, J.-C.; Clouet, E.; Leprêtre, F.; Bordas, E.; Arnal, B.

2018-06-01

Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20 MeV Kr8+ ions at 400 °C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)2 Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

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

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less

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

Goetsch, D.; Bieniussa, K.; Schulz, H.

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branchingmore » pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.« less

Metallurgical evaluation of a feedwater nozzle to safe-end weld

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

Bowerman, B.S.; Czajkowski, C.J.; Roberts, T.C.

1999-11-01

Weld cracks in safety class systems are a serious concern, because these systems are part of the primary barrier providing containment of radioactive coolant. Loss of weld integrity yields leaks, or, under catastrophic failure, can be the basis for a severe loss of coolant accident. A circumferential indication was found by ultrasonic examination (UT) in the N4A-2 inlet feedwater nozzle to safe-end weld during the second refueling outage of River Bend Station Unit 1 in March 1989. The indication, approximately 15cm (6in) long with a reported maximum depth of 0.5cm (0.1in), was located in the Alloy 182 weld butter onmore » the safe-end side of the weld. (The safe-end base metal was ASME SA 508 Class 1 carbon steel.) The reported characteristics of the UT indication were indicative of intergranular stress corrosion cracking. This indication was reexamined during the second and third fuel cycles in March 1990 and September 1991, respectively, and during the third refuel outage in November 1990. Crack growth was reported during each examination. The safe-end was replaced during the fourth refueling outage in the summer of 1992. The US Nuclear Regulatory Commission (NRC) subsequently contracted with Brookhaven National laboratory (BNL) to conduct a confirmatory investigation to establish the failure mode and determine the root causes of cracking in the safe-end weld.« less

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

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

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

Severe Accident Test Station Design Document

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

Snead, Mary A.; Yan, Yong; Howell, Michael

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure tomore » provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.« less

Chip breaking system for automated machine tool

DOEpatents

Arehart, Theodore A.; Carey, Donald O.

1987-01-01

The invention is a rotary selectively directional valve assembly for use in an automated turret lathe for directing a stream of high pressure liquid machining coolant to the interface of a machine tool and workpiece for breaking up ribbon-shaped chips during the formation thereof so as to inhibit scratching or other marring of the machined surfaces by these ribbon-shaped chips. The valve assembly is provided by a manifold arrangement having a plurality of circumferentially spaced apart ports each coupled to a machine tool. The manifold is rotatable with the turret when the turret is positioned for alignment of a machine tool in a machining relationship with the workpiece. The manifold is connected to a non-rotational header having a single passageway therethrough which conveys the high pressure coolant to only the port in the manifold which is in registry with the tool disposed in a working relationship with the workpiece. To position the machine tools the turret is rotated and one of the tools is placed in a material-removing relationship of the workpiece. The passageway in the header and one of the ports in the manifold arrangement are then automatically aligned to supply the machining coolant to the machine tool workpiece interface for breaking up of the chips as well as cooling the tool and workpiece during the machining operation.

Design and Testing for a New Thermosyphon Irradiation Vehicle

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

Felde, David K.; Carbajo, Juan J.; McDuffee, Joel Lee

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heatmore » loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.« less

Calculated effects of turbine rotor-blade cooling-air flow, altitude, and compressor bleed point on performance of a turbojet engine

NASA Technical Reports Server (NTRS)

Arne, Vernon L; Nachtigall, Alfred J

1951-01-01

Effects of air-cooling turbine rotor blades on performance of a turbojet engine were calculated for a range of altitudes from sea level to 40,000 feet and a range of coolant flows up to 3 percent of compressor air flow, for two conditions of coolant bleed from the compressor. Bleeding at required coolant pressure resulted in a sea-level thrust reduction approximately twice the percentage coolant flow and in an increase in specific fuel consumption approximately equal to percentage coolant flow. For any fixed value of coolant flow ratio the percentage thrust reduction and percentage increase in specific fuel consumption decreased with altitude. Bleeding coolant at the compressor discharge resulted in an additional 1 percent loss in performance at sea level and in smaller increase in loss of performance at higher altitudes.

Instrumentation Performance During the TMI-2 Accident

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.

2014-08-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. The loss of coolant and the hydrogen combustion that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

APT Blanket Thermal Analyses of Top Horizontal Row 1 Modules

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

Shadday, M.A.

1999-09-20

The Accelerator Production of Tritium (APT) cavity flood system (CFS) is designed to be the primary safeguard for the integrity of the blanket modules during loss of coolant accidents (LOCAs). For certain large break LOCAs the CFS also provides backup for the residual heat removal systems (RHRs) in cooling the target assemblies. In the unlikely event that the internal flow passages in a blanket module or target assembly dryout, decay heat in the metal structures will be dissipated to the CFS through the module or assembly walls (i.e., rung outer walls). The target assemblies consist of tungsten targets encased inmore » steel conduits, and they can safely sustain high metal temperatures. Under internally dry conditions, the cavity flood fluid will cool the target assemblies with vigorous nucleate boiling on the external surfaces. However, the metal structures in the blanket modules consist of lead cladded in aluminum, and they have a long-term exposure temperature limit currently set to 150 degrees C. Simultaneous LOCAs in both the target and blanket heat removal systems (HRS) could result in dryout of the target ladders, as well as the horizontal blanket modules above the target. The cavity flood coolant would boil on the outside surfaces of the target ladder rungs, and the resultant steam could reduce the effectiveness of convection heat transfer from the blanket modules to the cavity flood coolant. A two-part analysis was conducted to ascertain if the cavity flood system can adequately cool the blanket modules above the targets, even when boiling is occurring on the outer surfaces of the target ladder rungs. The first part of the analysis was to model transient thermal conduction in the front top horizontal row 1 module (i.e. top horizontal modules nearest the incoming beam), while varying parametrically the convection heat transfer coefficient (htc) for the external surfaces exposed to the cavity flood flow. This part of the analysis demonstrated that the module could adequately conduct heat to the outer module surfaces, given reasonable values for the convection heat transfer coefficients. The second part of the analysis consisted of two-phase flow modeling of the natural circulation of the cavity flood fluid past the top modules. Slots in the top shield allow the cavity flood fluid to circulate. The required width for these slots, to prevent steam from backing up and blanketing the outer surfaces of the top modules, was determined.« less

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

Parameter study on the influence of prepressurization on PWR fuel rod behavior during normal operation and hypothetical LOCAs

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

Brzoska, B.; Depisch, F.; Fuchs, H.P.

To analyze the influence of prepressurization on fuel rod behavior, a parametric study has been performed that considers the effects of as-fabricated fuel rod internal prepressure on the normal operation and postulated loss-of-coolant accident (LOCA) rod behavior of a 1300-MW(electric) Kraftwerk Union (KWU) standard pressurized water reactor nuclear power plant. A variation of the prepressure in the range from 15 to 35 bars has only a slight influence on normal operation behavior. Considering the LOCA behavior, only a small temperature increase results from prepressure reduction, while the core-wide straining behavior is improved significantly. The KWU prepressurization takes both conditions intomore » account.« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

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

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Catalytic igniters and their use to ignite lean hydrogen-air mixtures

DOEpatents

McLean, William J.; Thorne, Lawrence R.; Volponi, Joanne V.

1988-01-01

A catalytic igniter which can ignite a hydrogen-air mixture as lean as 5.5% hydrogen with induction times ranging from 20 s to 400 s, under conditions which may be present during a loss-of-liquid-coolant accident at a light water nuclear reactor comprises (a) a perforate catalytically active substrate, such as a platinum coated ceramic honeycomb or wire mesh screen, through which heated gases produced by oxidation of the mixture can freely flow and (b) a plurality of thin platinum wires mounted in a thermally conductive manner on the substrate and positioned thereon so as to be able to receive heat from the substrate and the heated gases while also in contact with unoxidized gases.

Atmospheric transport of radioactive debris to Norway in case of a hypothetical accident related to the recovery of the Russian submarine K-27.

PubMed

Bartnicki, Jerzy; Amundsen, Ingar; Brown, Justin; Hosseini, Ali; Hov, Øystein; Haakenstad, Hilde; Klein, Heiko; Lind, Ole Christian; Salbu, Brit; Szacinski Wendel, Cato C; Ytre-Eide, Martin Album

2016-01-01

The Russian nuclear submarine K-27 suffered a loss of coolant accident in 1968 and with nuclear fuel in both reactors it was scuttled in 1981 in the outer part of Stepovogo Bay located on the eastern coast of Novaya Zemlya. The inventory of spent nuclear fuel on board the submarine is of concern because it represents a potential source of radioactive contamination of the Kara Sea and a criticality accident with potential for long-range atmospheric transport of radioactive particles cannot be ruled out. To address these concerns and to provide a better basis for evaluating possible radiological impacts of potential releases in case a salvage operation is initiated, we assessed the atmospheric transport of radionuclides and deposition in Norway from a hypothetical criticality accident on board the K-27. To achieve this, a long term (33 years) meteorological database has been prepared and used for selection of the worst case meteorological scenarios for each of three selected locations of the potential accident. Next, the dispersion model SNAP was run with the source term for the worst-case accident scenario and selected meteorological scenarios. The results showed predictions to be very sensitive to the estimation of the source term for the worst-case accident and especially to the sizes and densities of released radioactive particles. The results indicated that a large area of Norway could be affected, but that the deposition in Northern Norway would be considerably higher than in other areas of the country. The simulations showed that deposition from the worst-case scenario of a hypothetical K-27 accident would be at least two orders of magnitude lower than the deposition observed in Norway following the Chernobyl accident. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

Comparison of the accident process, radioactivity release and ground contamination between Chernobyl and Fukushima-1.

PubMed

Imanaka, Tetsuji; Hayashi, Gohei; Endo, Satoru

2015-12-01

In this report, we have reviewed the basic features of the accident processes and radioactivity releases that occurred in the Chernobyl accident (1986) and in the Fukushima-1 accident (2011). The Chernobyl accident was a power-surge accident that was caused by a failure of control of a fission chain reaction, which instantaneously destroyed the reactor and building, whereas the Fukushima-1 accident was a loss-of-coolant accident in which the reactor cores of three units were melted by decay heat after losing the electricity supply. Although the quantity of radioactive noble gases released from Fukushima-1 exceeded the amount released from Chernobyl, the size of land area severely contaminated by (137)Cesium ((137)Cs) was 10 times smaller around Fukushima-1 compared with around Chernobyl. The differences in the accident process are reflected in the composition of the discharged radioactivity as well as in the composition of the ground contamination. Volatile radionuclides (such as (132)Te-(132)I, (131)I, (134)Cs and (137)Cs) contributed to the gamma-ray exposure from the ground contamination around Fukishima-1, whereas a greater variety of radionuclides contributed significantly around Chernobyl. When radioactivity deposition occurred, the radiation exposure rate near Chernobyl is estimated to have been 770 μGy h(-1) per initial (137)Cs deposition of 1000 kBq m(-2), whereas it was 100 μGy h(-1) around Fukushima-1. Estimates of the cumulative exposure for 30 years are 970 and 570 mGy per initial deposition of 1000 kBq m(-2) for Chernobyl and Fukusima-1, respectively. Of these exposures, 49 and 98% were contributed by radiocesiums ((134)Cs + (137)Cs) around Chernobyl and Fukushima-1, respectively. © The Author 2015. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

MARTINS: A foam/film flow model for molten material relocation in HWRs with U-Al-fueled multi-tube assemblies

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

Kalimullah

1994-03-01

Some special purpose heavy-water reactors (EM) are made of assemblies consisting of a number of coaxial aluminum-clad U-Al alloy fuel tubes and an outer Al sleeve surrounding the fuel tubes. The heavy water coolant flows in the annular gaps between the circular tubes. Analysis of severe accidents in such reactors requires a model for predicting the behavior of the fuel tubes as they melt and disrupt. This paper describes a detailed, mechanistic model for fuel tube heatup, melting, freezing, and molten material relocation, called MARTINS (Melting and Relocation of Tubes in Nuclear subassembly). The paper presents the modeling of themore » phenomena in MARTINS, and an application of the model to analysis of a reactivity insertion accident. Some models are being developed to compute gradual downward relocation of molten material at decay-heat power levels via candling along intact tubes, neglecting coolant vapor hydrodynamic forces on molten material. These models are inadequate for high power accident sequences involving significant hydrodynamic forces. These forces are included in MARTINS.« less

Comparison of Calculated and Experimental Temperatures and Coolant Pressure Losses for a Cascade of Small Air-Cooled Turbine Rotor Blades

NASA Technical Reports Server (NTRS)

Stepka, Francis S

1958-01-01

Average spanwise blade temperatures and cooling-air pressure losses through a small (1.4-in, span, 0.7-in, chord) air-cooled turbine blade were calculated and are compared with experimental nonrotating cascade data. Two methods of calculating the blade spanwise metal temperature distributions are presented. The method which considered the effect of the length-to-diameter ratio of the coolant passage on the blade-to-coolant heat-transfer coefficient and assumed constant coolant properties based on the coolant bulk temperature gave the best agreement with experimental data. The agreement obtained was within 3 percent at the midspan and tip regions of the blade. At the root region of the blade, the agreement was within 3 percent for coolant flows within the turbulent flow regime and within 10 percent for coolant flows in the laminar regime. The calculated and measured cooling-air pressure losses through the blade agreed within 5 percent. Calculated spanwise blade temperatures for assumed turboprop engine operating conditions of 2000 F turbine-inlet gas temperature and flight conditions of 300 knots at a 30,000-foot altitude agreed well with those obtained by the extrapolation of correlated experimental data of a static cascade investigation of these blades.

Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

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

Garbett, K; Mendler, O J; Gardner, G C

In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faultsmore » and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated.« less

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

DOE PAGES

Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

2015-08-25

Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

Effects of rotation on coolant passage heat transfer. Volume 1: Coolant passages with smooth walls

NASA Technical Reports Server (NTRS)

Hajek, T. J.; Wagner, J. H.; Johnson, B. V.; Higgins, A. W.; Steuber, G. D.

1991-01-01

An experimental program was conducted to investigate heat transfer and pressure loss characteristics of rotating multipass passages, for configurations and dimensions typical of modern turbine blades. The immediate objective was the generation of a data base of heat transfer and pressure loss data required to develop heat transfer correlations and to assess computational fluid dynamic techniques for rotating coolant passages. Experiments were conducted in a smooth wall large scale heat transfer model.

Ion irradiation testing and characterization of FeCrAl candidate alloys

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

Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew

2014-10-29

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commerciallymore » available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.« less

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

Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

NASA Astrophysics Data System (ADS)

Alat, Ece; Motta, Arthur T.; Comstock, Robert J.; Partezana, Jonna M.; Wolfe, Douglas E.

2016-09-01

In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO® coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti1-xAlxN (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm2 weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO® which showed a weight gain of 40.2 mg/dm2. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance.

Public Response to a Near-Miss Nuclear Accident Scenario Varying in Causal Attributions and Outcome Uncertainty.

PubMed

Cui, Jinshu; Rosoff, Heather; John, Richard S

2018-05-01

Many studies have investigated public reactions to nuclear accidents. However, few studies focused on more common events when a serious accident could have happened but did not. This study evaluated public response (emotional, cognitive, and behavioral) over three phases of a near-miss nuclear accident. Simulating a loss-of-coolant accident (LOCA) scenario, we manipulated (1) attribution for the initial cause of the incident (software failure vs. cyber terrorist attack vs. earthquake), (2) attribution for halting the incident (fail-safe system design vs. an intervention by an individual expert vs. a chance coincidence), and (3) level of uncertainty (certain vs. uncertain) about risk of a future radiation leak after the LOCA is halted. A total of 773 respondents were sampled using a 3 × 3 × 2 between-subjects design. Results from both MANCOVA and structural equation modeling (SEM) indicate that respondents experienced more negative affect, perceived more risk, and expressed more avoidance behavioral intention when the near-miss event was initiated by an external attributed source (e.g., earthquake) compared to an internally attributed source (e.g., software failure). Similarly, respondents also indicated greater negative affect, perceived risk, and avoidance behavioral intentions when the future impact of the near-miss incident on people and the environment remained uncertain. Results from SEM analyses also suggested that negative affect predicted risk perception, and both predicted avoidance behavior. Affect, risk perception, and avoidance behavior demonstrated high stability (i.e., reliability) from one phase to the next. © 2017 Society for Risk Analysis.

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

NASA Astrophysics Data System (ADS)

Borisov, V. M.; Trofimov, V. N.; Sapozhkov, A. Yu.; Kuzmenko, V. A.; Mikhaylov, V. B.; Cherkovets, V. Ye.; Yakushkin, A. A.; Yakushin, V. L.; Dzhumayev, P. S.

2016-12-01

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature ( T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al2O3, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

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

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Evaluation of RANS and LES models for Natural Convection in High-Aspect-Ratio Parallel Plate Channels

NASA Astrophysics Data System (ADS)

Fradeneck, Austen; Kimber, Mark

2017-11-01

The present study evaluates the effectiveness of current RANS and LES models in simulating natural convection in high-aspect ratio parallel plate channels. The geometry under consideration is based on a simplification of the coolant and bypass channels in the very high-temperature gas reactor (VHTR). Two thermal conditions are considered, asymmetric and symmetric wall heating with an applied heat flux to match Rayleigh numbers experienced in the VHTR during a loss of flow accident (LOFA). RANS models are compared to analogous high-fidelity LES simulations. Preliminary results demonstrate the efficacy of the low-Reynolds number k- ɛ formulations and their enhancement to the standard form and Reynolds stress transport model in terms of calculating the turbulence production due to buoyancy and overall mean flow variables.

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.

Gamma thermometer based reactor core liquid level detector

DOEpatents

Burns, Thomas J.

1983-01-01

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Nuclear Fuels & Materials Spotlight Volume 5

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

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

Numerical prediction of micro-channel LD heat sink operated with antifreeze based on CFD method

NASA Astrophysics Data System (ADS)

Liu, Gang; Liu, Yang; Wang, Chao; Wang, Wentao; Wang, Gang; Tang, Xiaojun

2014-12-01

To theoretically study the feasibility of antifreeze coolants applied as cooling fluids for high power LD heat sink, detailed Computational Fluid Dynamics (CFD) analysis of liquid cooled micro-channels heat sinks is presented. The performance operated with antifreeze coolant (ethylene glycol aqueous solution) compared with pure water are numerical calculated for the heat sinks with the same micro-channels structures. The maximum thermal resistance, total pressure loss (flow resistance), thermal resistance vs. flow-rate, and pressure loss vs. flow-rate etc. characteristics are numerical calculated. The results indicate that the type and temperature of coolants plays an important role on the performance of heat sinks. The whole thermal resistance and pressure loss of heat sinks increase significantly with antifreeze coolants compared with pure water mainly due to its relatively lower thermal conductivity and higher fluid viscosity. The thermal resistance and pressure loss are functions of the flow rate and operation temperature. Increasing of the coolant flow rate can reduce the thermal resistance of heat sinks; meanwhile increase the pressure loss significantly. The thermal resistance tends to a limit with increasing flow rate, while the pressure loss tends to increase exponentially with increasing flow rate. Low operation temperature chiefly increases the pressure loss rather than thermal resistance due to the remarkable increasing of fluid viscosity. The actual working point of the cooling circulation system can be determined on the basis of the pressure drop vs. flow rate curve for the micro-channel heat sink and that for the circulation system. In the same system, if the type or/and temperature of the coolant is changed, the working point is accordingly influenced, that is, working flow rate and pressure is changed simultaneously, due to which the heat sink performance is influenced. According to the numerical simulation results, if ethylene glycol aqueous solution is applied instead of pure water as the coolant under the same or a higher working temperature, the available output of optical power will decrease due to the worse heat sink performance; if applied under a lower working temperature(0 °C, －20 °C), although the heat sink performance become worse, however the temperature difference of heat transfer rises more significantly, the available output of optical power will increase on the contrary.

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

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

Borisov, V. M., E-mail: borisov@triniti.ru; Trofimov, V. N.; Sapozhkov, A. Yu.

2016-12-15

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature (T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al{sub 2}O{sub 3}, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppressionmore » of corrosion in liquid lead to the temperature of 720°C are shown.« less

Multiloop integral system test (MIST): Final report

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

Gloudemans, J.R.

1991-04-01

The Multiloop Integral System Test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility --more » the Once Through Integral System (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST program is reported in 11 volumes. Volumes 2 through 8 pertain to groups of Phase 3 tests by type; Volume 9 presents inter-group comparisons; Volume 10 provides comparisons between the RELAP5/MOD2 calculations and MIST observations, and Volume 11 (with addendum) presents the later Phase 4 tests. This is Volume 1 of the MIST final report, a summary of the entire MIST program. Major topics include, Test Advisory Group (TAG) issues, facility scaling and design, test matrix, observations, comparison of RELAP5 calculations to MIST observations, and MIST versus the TAG issues. MIST generated consistent integral-system data covering a wide range of transient interactions. MIST provided insight into integral system behavior and assisted the code effort. The MIST observations addressed each of the TAG issues. 11 refs., 29 figs., 9 tabs.« less

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

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

Karb, E.H.

1980-01-01

In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, whichmore » has been observed in all tests with previously irradiated rods, needs further investigation.« less

BWR zero pressure containment

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

Dillmann, C.W.; Townsend, H.E.; Nesbitt, L.B.

1992-02-25

This patent describes the operation of a nuclear reactor system, the system including a containment defining a drywall space wherein a nuclear reactor is disposed, there being a suppression pool in the containment with the suppression pool having a wetwell space above a level of the pool to which an non-condensable gases entering the suppression pool can vent. It comprises: continuously exhausting the wetwell space to remove gas mixture therefrom while admitting inflow of air from an atmospheric source thereof to the wetwell during normal operation by blocking off the inflow during a loss-of-coolant-accident whenever a pressure in the wetwellmore » space is above a predetermined value, and subjecting the gas subsequent to its removal from the wetwell to a treatment operation to separate any particulate material entrained therein from the gas mixture.« less

EMERALD REV.1. PWR Accident Activity Release

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

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1975-10-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« 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

Pressure suppression system

DOEpatents

Gluntz, D.M.

1994-10-04

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

Pressure suppression system

DOEpatents

Gluntz, Douglas M.

1994-01-01

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Interim Reliability Evaluation Program: analysis of the Browns Ferry, Unit 1, nuclear plant. Main report

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

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.

1982-07-01

A probabilistic risk assessment (PRA) was made of the Browns Ferry, Unit 1, nuclear plant as part of the Nuclear Regulatory Commission's Interim Reliability Evaluation Program (IREP). Specific goals of the study were to identify the dominant contributors to core melt, develop a foundation for more extensive use of PRA methods, expand the cadre of experienced PRA practitioners, and apply procedures for extension of IREP analyses to other domestic light water reactors. Event tree and fault tree analyses were used to estimate the frequency of accident sequences initiated by transients and loss of coolant accidents. External events such as floods,more » fires, earthquakes, and sabotage were beyond the scope of this study and were, therefore, excluded. From these sequences, the dominant contributors to probable core melt frequency were chosen. Uncertainty and sensitivity analyses were performed on these sequences to better understand the limitations associated with the estimated sequence frequencies. Dominant sequences were grouped according to common containment failure modes and corresponding release categories on the basis of comparison with analyses of similar designs rather than on the basis of detailed plant-specific calculations.« less

Solar receiver protection means and method for loss of coolant flow

DOEpatents

Glasgow, L.E.

1980-11-24

An apparatus and method are disclosed for preventing a solar receiver utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver by a plurality of reflectors which rotate so that they direct solar energy to the receiver as the earth rotates. The apparatus disclosed includes a first storage tank for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank having an inlet through which the coolant can enter. The first and second storage tanks are in fluid communication with each other through the solar receiver. The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank through the solar receiver and into the second storage tank. Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks will be sufficient to maintain the coolant in the receiver below a predetermined upper temperature until the solar reflectors become defocused with respect to the solar receiver due to the earth's rotation.

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

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Molecular structures and thermodynamic properties of monohydrated gaseous iodine compounds: Modelling for severe accident simulation

NASA Astrophysics Data System (ADS)

Sudolská, Mária; Cantrel, Laurent; Budzák, Šimon; Černušák, Ivan

2014-03-01

Monohydrated complexes of iodine species (I, I2, HI, and HOI) have been studied by correlated ab initio calculations. The standard enthalpies of formation, Gibbs free energy and the temperature dependence of the heat capacities at constant pressure were calculated. The values obtained have been implemented in ASTEC nuclear accident simulation software to check the thermodynamic stability of hydrated iodine compounds in the reactor coolant system and in the nuclear containment building of a pressurised water reactor during a severe accident. It can be concluded that iodine complexes are thermodynamically unstable by means of positive Gibbs free energies and would be represented by trace level concentrations in severe accident conditions; thus it is well justified to only consider pure iodine species and not hydrated forms.

Effects of turn region treatments on pressure loss through sharp 180-degree bends

NASA Astrophysics Data System (ADS)

Plevich, C. W.; Metzger, D. E.

An experimental study was conducted to evaluate the effect of geometric turn region inserts on pressure losses for flow through sharp 180-degree channel turns typical of internal cooling passages in gas turbine engine airfoils. The experiments were conducted in a rectangular cross-sectioned channel with 90-degree transverse rib roughening in both inlet and outlet legs, starting with completely smooth turn regions and progressing through various modifications including corner fillets, radial ribs, and turning vanes. The results show that modifications to the turn region geometry, particularly the inclusion of a single semi-circular turning vane, significantly reduce the pressure losses associated with coolant flows through sharp 180-degree turns and therefore can result in increased coolant flow for a given coolant supply pressure.

Aerodynamic losses calculation of a turbine blade with film cooling with forward and backward injection by numerical method

NASA Astrophysics Data System (ADS)

Prajapati, Anil

Thermal efficiency and power output of gas turbines can be increased by increasing the turbine blade inlet temperature. However, the main problem is the durability of the turbine blade due to the thermal stress on it at high temperature. This has led to the development of film cooling technology, in which coolant is injected from a series of cooling holes made on the blade surface to form an insulating blanket over the blade surface. However, it has to pay the aerodynamic penalties due to the injection of coolant, which are not fully understood. Pressure loss coefficient is one of the easy and widely used parameters to determine the aerodynamic loss occurred on a turbine blade. The losses occurred on the turbine blade with forward injection and backward injection cooling are studied at a different blowing ratios by a numerical simulation, which shows that the loss is higher in the case of backward injection than in forward injection. Fan-shaped cooling holes are also considered to compare with the cylindrical holes. It is observed that the loss is increased due to the fan-shaped holes in the forward injection whereas there is not a substantial difference due to the fan-shaped holes in the backward injection. The aerodynamic loss due to the location of coolant injection is studied by using injection from the leading edge, pressure side, suction side and trailing edge respectively. The study is performed to determine the effect of incidence angles and coolant injection angles on the aerodynamic loss.

Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

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

Gamble, Kyle A.; Hales, Jason D.

2016-12-01

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of themore » concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.« less

Downflow dryout in a heated ribbed vertical annulus with a cosine power profile (Results from test series ECS-2, WSR, and ECS-2cE)

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

Larson, T.K.; Anderson, J.L.; Condie, K.G.

Experiments designed to investigate surface dryout in a heated, ribbed annulus test section simulating one of the annular coolant channels of a Savannah River Plant production reactor Mark 22 fuel assembly have been conducted at the Idaho National Engineering Laboratory. The inner surface of the annulus was constructed of aluminum and was electrically heated to provide an axial cosine power profile and a flat azimuthal power shape. Data presented in this report are from the ECS-2, WSR, and ECS-2cE series of tests. These experiments were conducted to examine the onset of wall thermal excursion for a range of flow, inletmore » fluid temperature, and annulus outlet pressure. Hydraulic boundary conditions on the test section represent flowrates (0.1--1.4 1/s), inlet fluid temperatures (293--345 K), and outlet pressures (-18--139.7 cm of water relative to the bottom of the heated length (61--200 cm of water relative to the bottom of the lower plenum)) expected to occur during the Emergency Coolant System (ECS) phase of postulated Loss-of-Coolant Accident in a production reactor. The onset of thermal excursion based on the present data is consistent with data gathered in test rigs with flat axial power profiles. The data indicate that wall dryout is primarily a function of liquid superficial velocity. Air entrainment rate was observed to be a strong function of the boundary conditions (primarily flowrate and liquid temperature), but had a minor effect on the power at the onset of thermal excursion for the range of conditions examined. 14 refs., 33 figs., 13 tabs.« less

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

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

De Rosa, Felice

2006-07-01

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to themore » accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1. (author)« less

ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

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

2013-11-01

1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69more » rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in all respects except that it contained a partial blockage formed by attaching sleeves (or "balloons") to some of the rods. 6. SOURCE AND SCOPE OF DATA Phenomena Tested - Heat transfer in the core of a PWR during a re-flood phase of postulated large break LOCA. Test Designation - Achilles Rig. The programme includes the following types of experiments: - on an unballooned cluster: -- single phase air flow -- low pressure level swell -- low flooding rate re-flood -- high flooding rate re-flood - on a ballooned cluster containing 80% blockage formed by 16 balloon sleeves -- single phase air flow -- low flooding rate re-flood 7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM N/A 8. DATA FORMAT AND COMPUTER Many Computers (M00019MNYCP00). 9. TYPICAL RUNNING TIME N/A 11. CONTENTS OF LIBRARY The ACHILLES package contains test data and associated data processing software as well as the documentation listed above. 12. DATE OF ABSTRACT November 2013. KEYWORDS: DATABASES, BENCHMARKS, HEAT TRANSFER, LOSS-OF-COLLANT ACCIDENT, PWR REACTORS, REFLOODING« less

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

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

McHugh, Kevin M; Garnier, John E; Sergey Rashkeev

2013-01-01

Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a ß-SiC CMC overbraid, and sintered a-SiC weremore » tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.« less

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

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

Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M

2014-01-01

The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making themore » gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.« less

A Heated Tube Facility for Rocket Coolant Channel Research

NASA Technical Reports Server (NTRS)

Green, James M.; Pease, Gary M.; Meyer, Michael L.

1995-01-01

The capabilities of a heated tube facility used for testing rocket engine coolant channels at the NASA Lewis Research Center are presented. The facility uses high current, low voltage power supplies to resistively heat a test section to outer wall temperatures as high as 730 C (1350 F). Liquid or gaseous nitrogen, gaseous helium, or combustible liquids can be used as the test section coolant. The test section is enclosed in a vacuum chamber to minimize heat loss to the surrounding system. Test section geometry, size, and material; coolant properties; and heating levels can be varied to generate heat transfer and coolant performance data bases.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

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

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

System Study: Residual Heat Removal 1998-2014

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

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trends were identified in themore » RHR results. A highly statistically significant decreasing trend was observed for the RHR injection mode start-only unreliability. Statistically significant decreasing trends were observed for RHR shutdown cooling mode start-only unreliability and RHR shutdown cooling model 24-hour unreliability.« less

Integral isolation valve systems for loss of coolant accident protection

DOEpatents

Kanuch, David J.; DiFilipo, Paul P.

2018-03-20

A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

Passive containment cooling system

DOEpatents

Billig, P.F.; Cooke, F.E.; Fitch, J.R.

1994-01-25

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA. 1 figure.

Passive containment cooling system

DOEpatents

Billig, Paul F.; Cooke, Franklin E.; Fitch, James R.

1994-01-01

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA.

Pressure suppression containment system

DOEpatents

Gluntz, Douglas M.; Townsend, Harold E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Pressure suppression containment system

DOEpatents

Gluntz, D.M.; Townsend, H.E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

Aerodynamic effect of a honeycomb rotor tip shroud on a 50.8-centimeter-tip-diameter core turbine

NASA Technical Reports Server (NTRS)

Moffitt, T. P.; Whitney, W. J.

1983-01-01

A 50.8-cm-tip-diameter turbine equipped with a rotor tip shroud of hexagonal cell (or honeycomb) cross section has been tested in warm air (416 K) for a range of shroud coolant to primary flow rates. Test results were also obtained for the same turbine operated with a solid shroud for comparison. The results showed that the combined effect of the honeycomb shroud and the coolant flow was to cause a reduction of 2.8 points in efficiency at design speed, pressure ratio, and coolant flow rate. With the coolant system inactivated, the honeycomb shroud caused a decrease in efficiency of 2.3 points. These results and those obtained from a small reference turbine indicate that the dominant factor governing honeycomb tip shroud loss is the ratio of honeycomb depth to blade span. The loss results of the two shrouds could be correlated on this basis. The same honeycomb and coolant effects are expected to occur for the hot (2200 K) version of this turbine.

Combination pipe-rupture mitigator and in-vessel core catcher. [LMFBR

DOEpatents

Tilbrook, R.W.; Markowski, F.J.

1982-03-09

A device is described which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Combination pipe rupture mitigator and in-vessel core catcher

DOEpatents

Tilbrook, Roger W.; Markowski, Franz J.

1983-01-01

A device which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Solar receiver protection means and method for loss of coolant flow

DOEpatents

Glasgow, Lyle E.

1983-01-01

An apparatus and method for preventing a solar receiver (12) utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver (12) by a plurality of reflectors (16) which rotate so that they direct solar energy to the receiver (12) as the earth rotates. The apparatus disclosed includes a first storage tank (30) for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank (30) includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank (34) for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank (34) having an inlet through which the coolant can enter. The first and second storage tanks (30) and (34) are in fluid communication with each other through the solar receiver (12). The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank (30) through the solar receiver (12) and into the second storage tank (34). Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors (16) stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks (30) and (34) will be sufficient to maintain the coolant in the receiver (12) below a predetermined upper temperature until the solar reflectors (16) become defocused with respect to the solar receiver (12) due to the earth's rotation.

EMERALD REVISION 1; PWR accident activity release. [IBM360,370; FORTRAN IV

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

Fowler, T.B.; Tobias, M.L.; Fox, J.N.

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.IBM360,370; FORTRAN IV; OS/360,370 (IBM360,370); 520K bytes of memory are required..« less

10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

Code of Federal Regulations, 2014 CFR

2014-01-01

... following design basis events to ensure— (A) The integrity of the reactor coolant pressure boundary; (B) The...) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be...

10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

Code of Federal Regulations, 2013 CFR

2013-01-01

... following design basis events to ensure— (A) The integrity of the reactor coolant pressure boundary; (B) The...) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be...

10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

Code of Federal Regulations, 2010 CFR

2010-01-01

... following design basis events to ensure— (A) The integrity of the reactor coolant pressure boundary; (B) The...) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be...

46 CFR 196.30-20 - Breaking of safety valve seal.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 46 Shipping 7 2011-10-01 2011-10-01 false Breaking of safety valve seal. 196.30-20 Section 196.30... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-20 Breaking of safety valve seal. (a) If at any time it is necessary to break the seal on a safety valve for any purpose, the Chief...

46 CFR 196.30-20 - Breaking of safety valve seal.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Breaking of safety valve seal. 196.30-20 Section 196.30... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-20 Breaking of safety valve seal. (a) If at any time it is necessary to break the seal on a safety valve for any purpose, the Chief...

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).

Future Integrated Systems Concept for Preventing Aircraft Loss-of-Control Accidents

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.; Jacobson, Steven r.

2010-01-01

Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper presents future system concepts and research directions for preventing aircraft loss-of-control accidents.

Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

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

Pint, Bruce A.; Unocic, Kinga A.; Terrani, Kurt A.

2015-08-01

Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 [1-3] and this facility continues to support those efforts in the ATF community. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heatmore » and hydrogen generation in the core during a coolant-limited severe accident [4-5]. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.« less

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

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

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

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

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

Columbia Accident Investigation Board Report. Volume 1

NASA Technical Reports Server (NTRS)

Gehman, Harold W., Jr.; Barry, John L.; Deal, Duane W.; Hallock, James N.; Hess, Kenneth W.; Hubbard, G. Scott; Logsdon, John M.; Osheroff, Douglas D.; Ride, Sally K.; Tetrault, Roger E.

2003-01-01

The Columbia Accident Investigation Board's independent investigation into the tragic February 1, 2003, loss of the Space Shuttle Columbia and its seven-member crew lasted nearly seven months and involved 13 Board members, approximately 120 Board investigators, and thousands of NASA and support personnel. Because the events that initiated the accident were not apparent for some time, the investigation's depth and breadth were unprecedented in NASA history. Further, the Board determined early in the investigation that it intended to put this accident into context. We considered it unlikely that the accident was a random event; rather, it was likely related in some degree to NASA's budgets, history, and program culture, as well as to the politics, compromises, and changing priorities of the democratic process. We are convinced that the management practices overseeing the Space Shuttle Program were as much a cause of the accident as the foam that struck the left wing. The Board was also influenced by discussions with members of Congress, who suggested that this nation needed a broad examination of NASA's Human Space Flight Program, rather than just an investigation into what physical fault caused Columbia to break up during re-entry. Findings and recommendations are in the relevant chapters and all recommendations are compiled in Chapter 11. Volume I is organized into four parts: The Accident; Why the Accident Occurred; A Look Ahead; and various appendices. To put this accident in context, Parts One and Two begin with histories, after which the accident is described and then analyzed, leading to findings and recommendations. Part Three contains the Board's views on what is needed to improve the safety of our voyage into space. Part Four is reference material. In addition to this first volume, there will be subsequent volumes that contain technical reports generated by the Columbia Accident Investigation Board and NASA, as well as volumes containing reference documentation and other related material.

Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2

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

White, J.R.; Grush, W.H.; Keeler, C.D.

A preliminary posttest analysis of Loss-of-Coolant Experiment (LOCE) L2-2, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed to gain an understanding of the cause of the disparity between predicted and measured fuel rod cladding temperature responses in the LOFT core. LOCE L2-2 is the first experiment in the LOFT Power Ascension Series L2 (first series of LOFT nuclear experiments), which was designed to investigate the response of the LOFT nuclear core to the blowdown, refill, and reflood transients during LOCEs conducted at gradually increasing power levels. LOCE L2-2 was conducted at 50% power (25 MW, 26.38 kW/m).more » Results show that a core-wide rewet occurred early in the transient (during blowdown starting at about 7 s after rupture) which was not calculated in the pretest prediction analysis. This early core-wide rewet resulted in the peak fuel rod cladding temperatures being lower (by a mean value of 166/sup 0/K for 24 thermocouples) than had been calculated. This preliminary posttest analysis was concerned solely with determining why the early core-wide rewet was not predicted by the RELAP4/MOD6 pretest analysis and be no means is it a complete posttest analysis of LOCE L2-2 results. However, during this analysis, several errors made in the prettest analysis were found, and their impact on the predicted results is assessed. Three factors were postulated to have caused the disparity between predicted and measured fuel rod cladding temperatures for LOCE L2-2: (a) the initial fuel rod stored energy, (b) the heat transfer surface, and (c) the hydraulics calculation. These factors were examined and are discussed in this report. It was determined that core hydraulics, as influenced by the calculation of broken loop cold leg break flow, was the major factor causing the disparity.« less

Chemical Safety Alert: Catastrophic Failure of Storage Tanks

EPA Pesticide Factsheets

Aboveground, atmospheric storage tanks can fail when flammable vapors in the tank explode and break either the shell-to-bottom or side seam, resulting in hazardous release accidents. Proper maintenance practices can help prevent accidents.

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

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

Berry, Jan; Ferrada, Juan J; Curd, Warren

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predictedmore » to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.« less

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

Aircraft Loss-of-Control: Analysis and Requirements for Future Safety-Critical Systems and Their Validation

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.

2011-01-01

Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex, resulting from numerous causal and contributing factors acting alone or more often in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper summarizes recent analysis results in identifying worst-case combinations of loss-of-control accident precursors and their time sequences, a holistic approach to preventing loss-of-control accidents in the future, and key requirements for validating the associated technologies.

EMERALD REV. 1

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

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1974-03-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Concussion in Motor Vehicle Accidents: The Concussion Identification Index

ClinicalTrials.gov

2016-08-03

Motor Vehicle Accidents; TBI (Traumatic Brain Injury); Brain Contusion; Brain Injuries; Cortical Contusion; Concussion Mild; Cerebral Concussion; Brain Concussion; Accidents, Traffic; Traffic Accidents; Traumatic Brain Injury With Brief Loss of Consciousness; Traumatic Brain Injury With no Loss of Consciousness; Traumatic Brain Injury With Loss of Consciousness

Aircraft Loss-of-Control Accident Analysis

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.; Foster, John V.

2010-01-01

Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. To gain a better understanding into aircraft loss-of-control events and possible intervention strategies, this paper presents a detailed analysis of loss-of-control accident data (predominantly from Part 121), including worst case combinations of causal and contributing factors and their sequencing. Future potential risks are also considered.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

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

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

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

Aircraft Loss of Control Causal Factors and Mitigation Challenges

NASA Technical Reports Server (NTRS)

Jacobson, Steven R.

2010-01-01

Loss of control is the leading cause of jet fatalities worldwide. Aside from their frequency of occurrence, accidents resulting from loss of aircraft control seize the public s attention by yielding a large number of fatalities in a single event. In response to the rising threat to aviation safety, the NASA Aviation Safety Program has conducted a study of the loss of control problem. This study gathered four types of information pertaining to loss of control accidents: (1) statistical data; (2) individual accident reports that cite loss of control as a contributing factor; (3) previous meta-analyses of loss of control accidents; and (4) inputs solicited from aircraft manufacturers, air carriers, researchers, and other industry stakeholders. Using these information resources, the study team identified the causal factors that were cited in the greatest number of loss of control accidents, and which were emphasized most by industry stakeholders. This report describes the study approach, the key causal factors for aircraft loss of control, and recommended mitigation strategies to make near-term impacts, mid-term impacts, and Next Generation Air Transportation System impacts on the loss of control accident statistics

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

Cauquelin, C.

This paper presents an overview of the use of leak-before-break (LBB) analysis for EPR reactors. EPR is an evolutionary Nuclear Island of the 4 loop x 1500 Mwe class currently in the design phase. Application of LBB to the main coolant lines and resulting design impacts are summarized. Background information on LBB analysis in France and Germany is also presented.

19 CFR 125.35 - Report of loss, detention, or accident.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Report of loss, detention, or accident. 125.35 Section 125.35 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY..., detention, or accident. Any loss or detention of bonded merchandise, or any accident happening to a vehicle...

19 CFR 125.35 - Report of loss, detention, or accident.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 19 Customs Duties 1 2011-04-01 2011-04-01 false Report of loss, detention, or accident. 125.35 Section 125.35 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY..., detention, or accident. Any loss or detention of bonded merchandise, or any accident happening to a vehicle...

Opportunities for the LWR ATF materials development program to contribute to the LBE-cooled ADS materials qualification program

NASA Astrophysics Data System (ADS)

Gong, Xing; Li, Rui; Sun, Maozhou; Ren, Qisen; Liu, Tong; Short, Michael P.

2016-12-01

Accelerator-driven systems (ADS) are a promising approach for nuclear waste disposal. Nevertheless, the principal candidate materials proposed for ADS construction, such as the ferritic/martensitic steel, T91, and austenitic stainless steels, 316L and 15-15Ti, are not fully compatible with the liquid lead-bismuth eutectic (LBE) coolant. Under some operating conditions, liquid metal embrittlement (LME) or liquid metal corrosion (LMC) may occur in these steels when exposed to LBE. These environmentally-induced material degradation effects pose a threat to ADS reactor safety, as failure of the materials could initiate a severe accident, in which fission products are released into the coolant. Meanwhile, parallel efforts to develop accident-tolerant fuels (ATF) in light water reactors (LWRs) could provide both general materials design philosophies and specific material solutions to the ADS program. In this paper, the potential contributions of the ATF materials development program to the ADS materials qualification program are evaluated and discussed in terms of service conditions and materials performance requirements. Several specific areas where coordinated development may benefit both programs, including composite materials and selected coatings, are discussed.

Pretest and posttest calculations of Semiscale Test S-07-10D with the TRAC computer program. [PWR

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

Duerre, K.H.; Cort, G.E.; Knight, T.D.

The Transient Reactor Analysis Code (TRAC) developed at the Los Alamos National Laboratory was used to predict the behavior of the small-break experiment designated Semiscale S-07-10D. This test simulates a 10 per cent communicative cold-leg break with delayed Emergency Core Coolant injection and blowdown of the broken-loop steam generator secondary. Both pretest calculations that incorporated measured initial conditions and posttest calculations that incorporated measured initial conditions and measured transient boundary conditions were completed. The posttest calculated parameters were generally between those obtained from pretest calculations and those from the test data. The results are strongly dependent on depressurization rate and,more » hence, on break flow.« less

49 CFR 655.44 - Post-accident testing.

Code of Federal Regulations, 2010 CFR

2010-10-01

... practicable following an accident involving the loss of human life, an employer shall conduct drug and alcohol... accidents. (i) As soon as practicable following an accident not involving the loss of human life in which a...

Oxidation of 304 stainless steel in high-temperature steam

NASA Astrophysics Data System (ADS)

Ishida, Toshihisa; Harayama, Yasuo; Yaguchi, Sinnosuke

1986-08-01

An experiment on oxidation of 304 stainless steel was performed in steam between 900°C and 1350°C, using the spare cladding of the reactor of the nuclear-powered ship Mutsu. The temperature range was appropriate for a postulated loss of coolant accident (LOCA) analysis of a LWR. The oxidation kinetics were found to obey the parabolic law during the first period of 8 min. After the first period, the parabolic reaction rate constant decreased in the case of heating temperatures between 1100°C and 1250°C. At 1250°C, especially, a marked decrease was observed in the oxide scale-forming kinetics when the surface treated initially by mechanical polishing and given a residual stress. This enhanced oxidation resistance was attributed to the presence of a chromium-enriched layer which was detected by use of an X-ray microanalyzer. The oxidation kinetics equation obtained for the first 8 min is applicable to the model calculation of a hypothetical LOCA in a LWR, employing 304 stainless steel cladding.

Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

NASA Astrophysics Data System (ADS)

Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

2016-04-01

Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

Trace Assessment for BWR ATWS Analysis

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

Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

2010-04-22

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

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.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

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.

Economic Estimation of the Losses Caused by Surface Water Pollution Accidents in China From the Perspective of Water Bodies' Functions.

PubMed

Yao, Hong; You, Zhen; Liu, Bo

2016-01-22

The number of surface water pollution accidents (abbreviated as SWPAs) has increased substantially in China in recent years. Estimation of economic losses due to SWPAs has been one of the focuses in China and is mentioned many times in the Environmental Protection Law of China promulgated in 2014. From the perspective of water bodies' functions, pollution accident damages can be divided into eight types: damage to human health, water supply suspension, fishery, recreational functions, biological diversity, environmental property loss, the accident's origin and other indirect losses. In the valuation of damage to people's life, the procedure for compensation of traffic accidents in China was used. The functional replacement cost method was used in economic estimation of the losses due to water supply suspension and loss of water's recreational functions. Damage to biological diversity was estimated by recovery cost analysis and damage to environmental property losses were calculated using pollutant removal costs. As a case study, using the proposed calculation procedure the economic losses caused by the major Songhuajiang River pollution accident that happened in China in 2005 have been estimated at 2263 billion CNY. The estimated economic losses for real accidents can sometimes be influenced by social and political factors, such as data authenticity and accuracy. Besides, one or more aspects in the method might be overestimated, underrated or even ignored. The proposed procedure may be used by decision makers for the economic estimation of losses in SWPAs. Estimates of the economic losses of pollution accidents could help quantify potential costs associated with increased risk sources along lakes/rivers but more importantly, highlight the value of clean water to society as a whole.

Total Thermal Management of Battery Electric Vehicles (BEVs)

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

Lustbader, Jason A; Rugh, John P; Winkler, Jonathan M

The key hurdles to achieving wide consumer acceptance of battery electric vehicles (BEVs) are weather-dependent drive range, higher cost, and limited battery life. These translate into a strong need to reduce a significant energy drain and resulting drive range loss due to auxiliary electrical loads the predominant of which is the cabin thermal management load. Studies have shown that thermal subsystem loads can reduce the drive range by as much as 45% under ambient temperatures below -10 degrees C. Often, cabin heating relies purely on positive temperature coefficient (PTC) resistive heating, contributing to a significant range loss. Reducing this rangemore » loss may improve consumer acceptance of BEVs. The authors present a unified thermal management system (UTEMPRA) that satisfies diverse thermal and design needs of the auxiliary loads in BEVs. Demonstrated on a 2015 Fiat 500e BEV, this system integrates a semi-hermetic refrigeration loop with a coolant network and serves three functions: (1) heating and/or cooling vehicle traction components (battery, power electronics, and motor) (2) heating and cooling of the cabin, and (3) waste energy harvesting and re-use. The modes of operation allow a heat pump and air conditioning system to function without reversing the refrigeration cycle to improve thermal efficiency. The refrigeration loop consists of an electric compressor, a thermal expansion valve, a coolant-cooled condenser, and a chiller, the latter two exchanging heat with hot and cold coolant streams that may be directed to various components of the thermal system. The coolant-based heat distribution is adaptable and saves significant amounts of refrigerant per vehicle. Also, a coolant-based system reduces refrigerant emissions by requiring fewer refrigerant pipe joints. The authors present bench-level test data and simulation analysis and describe a preliminary control scheme for this system.« less

Reflux cooling experiments on the NCSU scaled PWR facility

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

Doster, J.M.; Giavedoni, E.

1993-01-01

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flowmore » rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.« less

Nuclear reactor flow control method and apparatus

DOEpatents

Church, J.P.

1993-03-30

Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Nuclear reactor flow control method and apparatus

DOEpatents

Church, John P.

1993-01-01

Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Probability of in-vessel steam explosion-induced containment failure for a KWU PWR

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

Esmaili, H.; Khatib-Rahbar, M.; Zuchuat, O.

During postulated core meltdown accidents in light water reactors, there is a likelihood for an in-vessel steam explosion when the melt contacts the coolant in the lower plenum. The objective of the work described in this paper is to determine the conditional probability of in-vessel steam explosion-induced containment failure for a Kraftwerk Union (KWU) pressurized water reactor (PWR). The energetics of the explosion depends on the mass of the molten fuel that mixes with the coolant and participates in the explosion and on the conversion of fuel thermal energy into mechanical work. The work can result in the generation ofmore » dynamic pressures that affect the lower head (and possibly lead to its failure), and it can cause acceleration of a slug (fuel and coolant material) upward that can affect the upper internal structures and vessel head and ultimately cause the failure of the upper head. If the upper head missile has sufficient energy, it can reach the containment shell and penetrate it. The analysis, must therefore, take into account all possible dissipation mechanisms.« less

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness ofmore » the PIUS concept to severe off-normal conditions having a very low probability of occurrence.« less

Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

NASA Astrophysics Data System (ADS)

Seguin, Nicolas

2014-05-01

In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical model, this numerical scheme is also efficient in terms of CPU time. Eventually, simpler models can locally replace the more complex model in order to simplify the overall computation, using some appropriate local error indicators developed in [5], without reducing the accuracy. References 1. Ishii, M., Hibiki, T., Thermo-fluid dynamics of two-phase flow, Springer, New-York, 2006. 2. Gallouët, T. and Hérard, J.-M., Seguin, N., Numerical modeling of two-phase flows using the two-fluid two-pressure approach, Math. Models Methods Appl. Sci., Vol. 14, 2004. 3. Seguin, N., Étude d'équations aux dérivées partielles hyperboliques en mécanique des fluides, Habilitation à diriger des recherches, UPMC-Paris 6, 2011. 4. Coquel, F., Hérard, J-M., Saleh, K., Seguin, N., A Robust Entropy-Satisfying Finite Volume Scheme for the Isentropic Baer-Nunziato Model, ESAIM: Mathematical Modelling and Numerical Analysis, Vol. 48, 2013. 5. Mathis, H., Cancès, C., Godlewski, E., Seguin, N., Dynamic model adaptation for multiscale simulation of hyperbolic systems with relaxation, preprint, 2013.

An Experimental and Numerical Investigation of Endwall Aerodynamics and Heat Transfer in a Gas Turbine Nozzle Guide Vane with Slot Film Cooling

NASA Astrophysics Data System (ADS)

Alqefl, Mahmood Hasan

In many regions of the high-pressure gas turbine, film cooling flows are used to protect the turbine components from the combustor exit hot gases. Endwalls are challenging to cool because of the complex system of secondary flows that disturb surface film coolant coverage. The secondary flow vortices wash the film coolant from the surface into the mainstream significantly decreasing cooling effectiveness. In addition to being effected by secondary flow structures, film cooling flow can also affect these structures by virtue of their momentum exchange. In addition, many studies in the literature have shown that endwall contouring affects the strength of passage secondary flows. Therefore, to develop better endwall cooling schemes, a good understanding of passage aerodynamics and heat transfer as affected by interactions of film cooling flows with secondary flows is required. This experimental and computational study presents results from a linear, stationary, two-passage cascade representing the first stage nozzle guide vane of a high-pressure gas turbine with an axisymmetrically contoured endwall. The sources of film cooling flows are upstream combustor liner coolant and endwall slot film coolant injected immediately upstream of the cascade passage inlet. The operating conditions simulate combustor exit flow features, with a high Reynolds number of 390,000 and approach flow turbulence intensity of 11% with an integral length scale of 21% of the chord length. Measurements are performed with varying slot film cooling mass flow to mainstream flow rate ratios (MFR). Aerodynamic effects are documented with five-hole probe measurements at the exit plane. Heat transfer is documented through recovery temperature measurements with a thermocouple. General secondary flow features are observed. Total pressure loss measurements show that varying the slot film cooling MFR has some effects on passage loss. Velocity vectors and vorticity distributions show a very thin, yet intense, cross-pitch flow on the contoured endwall side. Endwall adiabatic effectiveness values and coolant distribution thermal fields show minimal effects of varying slot film coolant MFR. This suggests the dominant effects of combustor liner coolant. show dominant effects of combustor liner coolant on cooling the endwall. A coolant vorticity correlation presenting the advective mixing of the coolant due to secondary flow vorticity at the exit plane is also discussed.

Aircraft Loss of Control Study

NASA Technical Reports Server (NTRS)

Jacobson, Steven R.

2010-01-01

Loss of control has become the leading cause of jet fatalities worldwide. Aside from their frequency of occurrence, accidents resulting from loss of aircraft control seize the public s attention by yielding large numbers of fatalities in a single event. In response to the rising threat to aviation safety, NASA's Aviation Safety Program has conducted a study of the loss of control problem. This study gathered four types of information pertaining to loss of control accidents: (1) statistical data; (2) individual accident reports that cite loss of control as a contributing factor; (3) previous meta-analyses of loss of control accidents; and (4) inputs solicited from aircraft manufacturers, air carriers, researchers, and other industry stakeholders. Using these information resources, the study team identified causal factors that were cited in the greatest number of loss of control accidents, and which were emphasized most by industry stakeholders. For each causal factor that was linked to loss of control, the team solicited ideas about what solutions are required and future research efforts that could potentially help avoid their occurrence or mitigate their consequences when they occurred in flight.

The willingness to pay of parties to traffic accidents for loss of productivity and consolation compensation.

PubMed

Jou, Rong-Chang; Chen, Tzu-Ying

2015-12-01

In this study, willingness to pay (WTP) for loss of productivity and consolation compensation by parties to traffic accidents is investigated using the Tobit model. In addition, WTP is compared to compensation determined by Taiwanese courts. The modelling results showed that variables such as education, average individual monthly income, traffic accident history, past experience of severe traffic accident injuries, the number of working days lost due to a traffic accident, past experience of accepting compensation for traffic accident-caused productivity loss and past experience of accepting consolation compensation caused by traffic accidents have a positive impact on WTP. In addition, average WTP for these two accident costs were obtained. We found that parties to traffic accidents were willing to pay more than 90% of the compensation determined by the court in the scenario of minor and moderate injuries. Parties were willing to pay approximately 80% of the compensation determined by the court for severe injuries, disability and fatality. Therefore, related agencies can use our study findings as the basis for determining the compensation that parties should pay for productivity losses caused by traffic accidents of different types. Copyright © 2015 Elsevier Ltd. All rights reserved.

Effect of mass transfer processes on accumulation and crystallization of boric acid in WWER core in emergency cases

NASA Astrophysics Data System (ADS)

Morozov, A. V.; Pityk, A. V.; Ragulin, S. V.; Sahipgareev, A. R.; Soshkina, A. S.; Shlepkin, A. S.

2017-09-01

In this paper the processes of boric acid mass transfer in a WWER-TOI nuclear reactor in case of the accidents with main coolant circuit rupture and operation of passive safety systems (the hydro accumulators systems of the first, second and third stages, as well as the passive heat removal system) are considered. The results of the calculation of changes in the boric acid solution concentration in the core for the WWER emergency mode are presented. According to the results of the calculation a significant excess of the ultimate concentration of boric acid in accidents with main coolant circuit rupture after 43 hours of emergency mode is observed. The positive influence of the boric acid droplet entrainment on the processes of its crystallization and accumulation in the core is shown. The mass of boric acid deposits on the internals is determined. The received results allow concluding that the accumulation and crystallization of boric acid in the core may lead to blocking the flow cross section and to deterioration of heat removal from fuel rods. The necessity of an experimental studies of the processes of boric acid drop entrainment under conditions specific to the WWER emergency modes is shown.

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Nuclear reactor melt arrest and coolability device

DOEpatents

Theofanous, Theo G.; Dinh, Nam Truc; Wachowiak, Richard M.

2016-06-14

Example embodiments provide a Basemat-Internal Melt Arrest and Coolability device (BiMAC) that offers improved spatial and mechanical characteristics for use in damage prevention and risk mitigation in accident scenarios. Example embodiments may include a BiMAC having an inclination of less than 10-degrees from the basemat floor and/or coolant channels of less than 4 inches in diameter, while maintaining minimum safety margins required by the Nuclear Regulatory Commission.

CFD Applications in Support of the Space Shuttle Risk Assessment

NASA Technical Reports Server (NTRS)

Baum, Joseph D.; Mestreau, Eric; Luo, Hong; Sharov, Dmitri; Fragola, Joseph; Loehner, Rainald; Cook, Steve (Technical Monitor)

2000-01-01

The paper describes a numerical study of a potential accident scenario of the space shuttle, operating at the same flight conditions as flight 51L, the Challenger accident. The interest in performing this simulation is derived by evidence that indicates that the event itself did not exert large enough blast loading on the shuttle to break it apart. Rather, the quasi-steady aerodynamic loading on the damaged, unbalance vehicle caused the break-up. Despite the enormous explosive potential of the shuttle total fuel load (both liquid and solid), the post accident explosives working group estimated the maximum energy involvement to be equivalent to about five hundreds of pounds of TNT. This understanding motivated the simulation described here. To err on the conservative side, we modeled the event as an explosion, and used the maximum energy estimate. We modeled the transient detonation of a 500 lbs spherical charge of TNT, placed at the main engine, and the resulting blast wave propagation about the complete stack. Tracking of peak pressures and impulses at hundreds of locations on the vehicle surface indicate that the blast load was insufficient to break the vehicle, hence demonstrating likely crew survivability through such an event.

Development of a cleaning process for uranium chips machined with a glycol-water-borax coolant

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

Taylor, P.A.

1984-12-01

A chip-cleaning process has been developed to remove the new glycol-water-borax coolant from oralloy chips. The process involves storing the freshly cut chips in Freon-TDF until they are cleaned, washing with water, and displacing the water with Freon-TDF. The wash water can be reused many times and still yield clean chips and then be added to the coolant to make up for evaporative losses. The Freon-TDF will be cycled by evaporation. The cleaning facility is currently being designed and should be operational by April 1985.

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

NASA Astrophysics Data System (ADS)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

Economic Estimation of the Losses Caused by Surface Water Pollution Accidents in China From the Perspective of Water Bodies’ Functions

PubMed Central

Yao, Hong; You, Zhen; Liu, Bo

2016-01-01

The number of surface water pollution accidents (abbreviated as SWPAs) has increased substantially in China in recent years. Estimation of economic losses due to SWPAs has been one of the focuses in China and is mentioned many times in the Environmental Protection Law of China promulgated in 2014. From the perspective of water bodies’ functions, pollution accident damages can be divided into eight types: damage to human health, water supply suspension, fishery, recreational functions, biological diversity, environmental property loss, the accident’s origin and other indirect losses. In the valuation of damage to people’s life, the procedure for compensation of traffic accidents in China was used. The functional replacement cost method was used in economic estimation of the losses due to water supply suspension and loss of water’s recreational functions. Damage to biological diversity was estimated by recovery cost analysis and damage to environmental property losses were calculated using pollutant removal costs. As a case study, using the proposed calculation procedure the economic losses caused by the major Songhuajiang River pollution accident that happened in China in 2005 have been estimated at 2263 billion CNY. The estimated economic losses for real accidents can sometimes be influenced by social and political factors, such as data authenticity and accuracy. Besides, one or more aspects in the method might be overestimated, underrated or even ignored. The proposed procedure may be used by decision makers for the economic estimation of losses in SWPAs. Estimates of the economic losses of pollution accidents could help quantify potential costs associated with increased risk sources along lakes/rivers but more importantly, highlight the value of clean water to society as a whole. PMID:26805869

Mountain Biking with Groups: A "Safe" Activity?

ERIC Educational Resources Information Center

Allen, Terry

2001-01-01

A survey mailed to 200 British mountain bike leaders found that rates of cycling accidents and injuries were greater in forests and woodlands than on terrain where a license is required to lead groups of young cyclists. Excessive speed was mentioned in most accidents, coupled with poor use of breaks in many cases. (SV)

Gas Retention, Gas Release, and Fluidization of Spherical Resorcinol-Formaldehyde (sRF) Ion Exchange Resin

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

Gauglitz, Phillip A.; Rassat, Scot D.; Linn, Diana

The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs in ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. When loaded with radioactive Cs, radiolysis of water in the LAW liquid will generate hydrogen gas. In normal operations, the generated hydrogen is expected to remainmore » dissolved in the liquid and be continuously removed by liquid flow. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin bed and below the bottom screen that supports the resin within the column, which creates a hydrogen flammability hazard. Because there is a potential for a large fraction of the retained hydrogen to be released over a short duration as a gas release event, there is a need to quantify the size and rate of potential gas release events. Due to the potential for a large, rapid gas release event, an evaluation of mitigation methods to eliminate the hydrogen hazard is also needed. One method being considered for mitigating the hydrogen hazard during a loss of flow accident is to have a secondary flow system, with two redundant pumps operating in series, that re-circulates liquid upwards through the bed and into a vented break tank where hydrogen gas is released from the liquid and removed by venting the headspace of the break tank. The mechanism for inducing release of gas from the sRF bed is to fluidize the bed, which should allow retained bubbles to rise and be carried to the break tank. The overall conclusion of the testing is that fluidization is an effective method to remove hydrogen gas from a bed of sRF resin, but that a single fluidization velocity that is adequate to release gas in 55 ºC water will over-fluidize sRF resin in most LAW liquids, including both nominal and high-limit LAW simulants used in testing. An upper packed bed can retain hydrogen gas and pose a flammability hazard. Using periodic on:off fluidization, such as 5:55 min. on:off cycles, is effective at releasing gas while not creating an upper packed bed. Note that lengthening the fluidization duration in a one-hour cycle did result in a stable upper packed bed in one case with the nominal LAW simulant, so testing focused on shorter “on” periods which are needed for effective hydrogen release with periodic on:off fluidization« less

Numerical Simulation of Hydrodynamics of a Heavy Liquid Drop Covered by Vapor Film in a Water Pool

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

Ma, W.M.; Yang, Z.L.; Giri, A.

2002-07-01

A numerical study on the hydrodynamics of a droplet covered by vapor film in water pool is carried out. Two level set functions are used as to implicitly capture the interfaces among three immiscible fluids (melt-drop, vapor and coolant). This approach leaves only one set of conservation equations for the three phases. A high-order Navier-Stokes solver, called Cubic-Interpolated Pseudo-Particle (CIP) algorithm, is employed in combination with level set approach, which allows large density ratios (up to 1000), surface tension and jump in viscosity. By this calculation, the hydrodynamic behavior of a melt droplet falling into a volatile coolant is simulated,more » which is of great significance to reveal the mechanism of steam explosion during a hypothetical severe reactor accident. (authors)« less

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for a loss of offsite power initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following amore » loss of offsite power.« less

Safety and health of professional drivers who drive on Brazilian highways

PubMed Central

Narciso, Fernanda Veruska; de Mello, Marco Túlio

2017-01-01

ABSTRACT Traffic accidents and resulting injuries and deaths have become a global epidemic. In Brazil, most professional drivers, especially truck drivers, face irregular working hours and can be awake for more than 18 hours/day, which reduces their performance and alertness. In this article, we discuss the laws related to Brazilian professional drivers and their current amendments (No. 12,619/2012 and No. 13,103/2015) in relation to working hours at the wheel and rest breaks, which are vital for the quality of life of drivers and society in general. We note that the new law appears to be less efficient than the previous one as it causes insecurity and concern to the users of the transportation system, drivers, and employers. To restrict and reduce accidents, deaths, and injuries in traffic, appropriate legislation is essential, aiming at the safety of workers and users of highways. The law must also benefit the commercial aspect, strengthening the reduction in production and logistics losses. Additionally, traffic education programs are needed, as well as better supervision in relation to total working hours. PMID:28380210

Safety and environmental aspects of organic coolants for fusion facilities

NASA Astrophysics Data System (ADS)

Natalizio, A.; Hollies, R. E.; Gierszewski, P.

1993-06-01

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350-400°C) at modest pressure (2-4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, J. R. (Inventor)

1985-01-01

A jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A TEMP, responsive to the heat from the coolant in the secondary flow path, automatically pumps the withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature. At this lower temperature, the TEMP/jet jump combination ceases its circulation boosting operation. The TEMP/jet pump combination is automatic, self-regulating and provides an emergency pumping system free of moving parts.

Uncertainty quantification for accident management using ACE surrogates

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

Varuttamaseni, A.; Lee, J. C.; Youngblood, R. W.

The alternating conditional expectation (ACE) regression method is used to generate RELAP5 surrogates which are then used to determine the distribution of the peak clad temperature (PCT) during the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed (F and B) operation in the Zion-1 nuclear power plant. The construction of the surrogates assumes conditional independence relations among key reactor parameters. The choice of parameters to model is based on the macroscopic balance statements governing the behavior of the reactor. The peak clad temperature is calculated based on the independent variables that are known tomore » be important in determining the success of the F and B operation. The relationship between these independent variables and the plant parameters such as coolant pressure and temperature is represented by surrogates that are constructed based on 45 RELAP5 cases. The time-dependent PCT for different values of F and B parameters is calculated by sampling the independent variables from their probability distributions and propagating the information through two layers of surrogates. The results of our analysis show that the ACE surrogates are able to satisfactorily reproduce the behavior of the plant parameters even though a quasi-static assumption is primarily used in their construction. The PCT is found to be lower in cases where the F and B operation is initiated, compared to the case without F and B, regardless of the F and B parameters used. (authors)« less

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

An Accident of History: Breaking the District Monopoly on Public School Facilities

ERIC Educational Resources Information Center

Smith, Nelson

2012-01-01

Traditional public school districts hold a monopoly over the financing and ownership of public education facilities. With rare exceptions, public charter schools have no legal claim to these buildings. This monopoly is an accident of history. It would never have developed had there been substantial numbers of other public schools, not supervised…

Final Report on ITER Task Agreement 81-08

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

Richard L. Moore

As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of themore » ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.« less

Precursors of dangerous substances formed in the loss of control of chemical systems.

PubMed

Cozzani, V; Zanelli, S

1999-03-01

Article 2 of Directive 96/82/EC on the control of major accident hazards caused by dangerous substances requires to consider also the hazards due to the dangerous substances "which it is believed may be generated during loss of control of an industrial chemical process", although no generally accepted guidelines are available for the identification of these substances. In the present study, the accidents involving the unwanted formation of dangerous substances as a consequence of the loss of control of chemical systems were investigated. A specifically developed database was used, containing data on more than 400 of these accidents and on the substances involved. The hazardous substances formed in the accidents and the precursors of these substances were identified. The influence of accident characteristics on the substances formed was investigated. In the context of the application of Directive 96/82/EC, an accident severity index and a hazard rating of the precursors of dangerous substances formed in the accidents were proposed. A lumping approach was used in order to develop schemes for the preliminary identification of substances that may be formed in the loss of control of chemical system. The results of accident analysis were used to test the schemes developed.

Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

NASA Astrophysics Data System (ADS)

Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

2014-05-01

The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

NASA Astrophysics Data System (ADS)

Narukawa, Takafumi; Yamaguchi, Akira; Jang, Sunghyon; Amaya, Masaki

2018-02-01

For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions of light-water-reactors, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best among the three models to estimate the fracture probability in terms of the degree of prediction accuracy for both next data to be obtained and the true model. Using the log-probit model, it was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, D.M.

1996-03-12

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Preparation of UC0.07-0.10N0.90-0.93 spheres for TRISO coated fuel particles

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, C. M.; Lindemer, T. B.; Johnson, J. A.; Collins, J. L.

2014-05-01

The US Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with dense uranium nitride (UN) kernels with diameters of 650 or 800 μm. The objectives of this effort are to make uranium oxide microspheres with adequately dispersed carbon nanoparticles and to convert these microspheres into UN spheres, which could be then sintered into kernels. Recent improvements to the internal gelation process were successfully applied to the production of uranium gel spheres with different concentrations of carbon black. After the spheres were washed and dried, a simple two-step heat profile was used to produce porous microspheres with a chemical composition of UC0.07-0.10N0.90-0.93. The first step involved heating the microspheres to 2023 K in a vacuum, and in the second step, the microspheres were held at 1873 K for 6 h in flowing nitrogen.

Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

NASA Astrophysics Data System (ADS)

Shriwastaw, R. S.; Sawarn, Tapan K.; Banerjee, Suparna; Rath, B. N.; Dubey, J. S.; Kumar, Sunil; Singh, J. L.; Bhasin, Vivek

2017-09-01

The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant-accident (LOCA) condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 °C at an interval of 50 °C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%.

Post Quench Ductility Evaluation of Zircaloy-4 and Select Iron Alloys under Design Basis and Extended LOCA Conditions

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

Yan, Yong; Keiser, James R; Terrani, Kurt A

2014-01-01

Oxidation experiments were conducted at 1200 C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 were determined to be k = 2.173 107 g2/cm4/s C, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post quenched samples was evaluated by ring compression tests atmore » 135 C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.« less

Velocity and void distribution in a counter-current two-phase flow

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

Gabriel, S.; Schulenberg, T.; Laurien, E.

2012-07-01

Different flow regimes were investigated in a horizontal channel. Simulating a hot leg injection in case of a loss of coolant accident or flow conditions in reflux condenser mode, the hydraulic jump and partially reversed flow were identified as major constraints for a high amount of entrained water. Trying to simulate the reflux condenser mode, the test section now includes an inclined section connected to a horizontal channel. The channel is 90 mm high and 110 mm wide. Tests were carried out for water and air at ambient pressure and temperature. High speed video-metry was applied to obtain velocities frommore » flow pattern maps of the rising and falling fluid. In the horizontal part of the channel with partially reversed flow the fluid velocities were measured by planar particle image velocimetry. To obtain reliable results for the gaseous phase, this analysis was extended by endoscope measurements. Additionally, a new method based on the optical refraction at the interface between air and water in a back-light was used to obtain time-averaged void fraction. (authors)« less

Evaluating the influential priority of the factors on insurance loss of public transit

PubMed Central

Su, Yongmin; Chen, Xinqiang

2018-01-01

Understanding correlation between influential factors and insurance losses is beneficial for insurers to accurately price and modify the bonus-malus system. Although there have been a certain number of achievements in insurance losses and claims modeling, limited efforts focus on exploring the relative role of accidents characteristics in insurance losses. The primary objective of this study is to evaluate the influential priority of transit accidents attributes, such as the time, location and type of accidents. Based on the dataset from Washington State Transit Insurance Pool (WSTIP) in USA, we implement several key algorithms to achieve the objectives. First, K-means algorithm contributes to cluster the insurance loss data into 6 intervals; second, Grey Relational Analysis (GCA) model is applied to calculate grey relational grades of the influential factors in each interval; in addition, we implement Naive Bayes model to compute the posterior probability of factors values falling in each interval. The results show that the time, location and type of accidents significantly influence the insurance loss in the first five intervals, but their grey relational grades show no significantly difference. In the last interval which represents the highest insurance loss, the grey relational grade of the time is significant higher than that of the location and type of accidents. For each value of the time and location, the insurance loss most likely falls in the first and second intervals which refers to the lower loss. However, for accidents between buses and non-motorized road users, the probability of insurance loss falling in the interval 6 tends to be highest. PMID:29298337

Evaluating the influential priority of the factors on insurance loss of public transit.

PubMed

Zhang, Wenhui; Su, Yongmin; Ke, Ruimin; Chen, Xinqiang

2018-01-01

Understanding correlation between influential factors and insurance losses is beneficial for insurers to accurately price and modify the bonus-malus system. Although there have been a certain number of achievements in insurance losses and claims modeling, limited efforts focus on exploring the relative role of accidents characteristics in insurance losses. The primary objective of this study is to evaluate the influential priority of transit accidents attributes, such as the time, location and type of accidents. Based on the dataset from Washington State Transit Insurance Pool (WSTIP) in USA, we implement several key algorithms to achieve the objectives. First, K-means algorithm contributes to cluster the insurance loss data into 6 intervals; second, Grey Relational Analysis (GCA) model is applied to calculate grey relational grades of the influential factors in each interval; in addition, we implement Naive Bayes model to compute the posterior probability of factors values falling in each interval. The results show that the time, location and type of accidents significantly influence the insurance loss in the first five intervals, but their grey relational grades show no significantly difference. In the last interval which represents the highest insurance loss, the grey relational grade of the time is significant higher than that of the location and type of accidents. For each value of the time and location, the insurance loss most likely falls in the first and second intervals which refers to the lower loss. However, for accidents between buses and non-motorized road users, the probability of insurance loss falling in the interval 6 tends to be highest.

High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station

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

Wong, S.; DiBiasio, A.; Gunther, W.

1993-09-01

The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failuremore » modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.« less

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

RELAP5 Application to Accident Analysis of the NIST Research Reactor

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

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

Overflow control valve

DOEpatents

Hundal, Rolv; Kessinger, Boyd A.; Parlak, Edward A.

1984-07-24

An overflow control valve for use in a liquid sodium coolant pump tank which valve can be extended to create a seal with the pump tank wall or retracted to break the seal thereby accommodating valve removal. An actuating shaft which controls valve disc position also has cams which bear on roller surfaces to force retraction of a sliding cylinder against spring tension to retract the cylinder from sealing contact with the pump tank.

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, James R. (Inventor)

1987-01-01

The invention does away with the necessity of moving parts such as a check valve in a nuclear reactor cooling system. Instead, a jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A main flow exists for a reactor coolant. A point of withdrawal is provided for a secondary flow. A TEMP, responsive to the heat from said coolant in the secondary flow path, automatically pumps said withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature where the heat is no longer a problem. At this lower temperature, the TEMP/jet pump combination ceases its circulation boosting operation. When the nuclear reactor is restarted and the coolant again exceeds the lower temperature setting, the TEMP/jet pump automatically resumes operation. The TEMP/jet pump combination is thus automatic, self-regulating and provides an emergency pumping system free of moving parts.

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

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

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

Traits and causes of environmental loss-related chemical accidents in China based on co-word analysis.

PubMed

Wu, Desheng; Song, Yu; Xie, Kefan; Zhang, Baofeng

2018-04-25

Chemical accidents are major causes of environmental losses and have been debated due to the potential threat to human beings and environment. Compared with the single statistical analysis, co-word analysis of chemical accidents illustrates significant traits at various levels and presents data into a visual network. This study utilizes a co-word analysis of the keywords extracted from the Web crawling texts of environmental loss-related chemical accidents and uses the Pearson's correlation coefficient to examine the internal attributes. To visualize the keywords of the accidents, this study carries out a multidimensional scaling analysis applying PROXSCAL and centrality identification. The research results show that an enormous environmental cost is exacted, especially given the expected environmental loss-related chemical accidents with geographical features. Meanwhile, each event often brings more than one environmental impact. Large number of chemical substances are released in the form of solid, liquid, and gas, leading to serious results. Eight clusters that represent the traits of these accidents are formed, including "leakage," "poisoning," "explosion," "pipeline crack," "river pollution," "dust pollution," "emission," and "industrial effluent." "Explosion" and "gas" possess a strong correlation with "poisoning," located at the center of visualization map.

The effects of aircraft certification rules on general aviation accidents

NASA Astrophysics Data System (ADS)

Anderson, Carolina Lenz

The purpose of this study was to analyze the frequency of general aviation airplane accidents and accident rates on the basis of aircraft certification to determine whether or not differences in aircraft certification rules had an influence on accidents. In addition, the narrative cause descriptions contained within the accident reports were analyzed to determine whether there were differences in the qualitative data for the different certification categories. The certification categories examined were: Federal Aviation Regulations Part 23, Civil Air Regulations 3, Light Sport Aircraft, and Experimental-Amateur Built. The accident causes examined were those classified as: Loss of Control, Controlled Flight into Terrain, Engine Failure, and Structural Failure. Airworthiness certification categories represent a wide diversity of government oversight. Part 23 rules have evolved from the initial set of simpler design standards and have progressed into a comprehensive and strict set of rules to address the safety issues of the more complex airplanes within the category. Experimental-Amateur Built airplanes have the least amount of government oversight and are the fastest growing segment. The Light Sport Aircraft category is a more recent certification category that utilizes consensus standards in the approval process. Civil Air Regulations 3 airplanes were designed and manufactured under simpler rules but modifying these airplanes has become lengthy and expensive. The study was conducted using a mixed methods methodology which involves both quantitative and qualitative elements. A Chi-Square test was used for a quantitative analysis of the accident frequency among aircraft certification categories. Accident rate analysis of the accidents among aircraft certification categories involved an ANCOVA test. The qualitative component involved the use of text mining techniques for the analysis of the narrative cause descriptions contained within the accident reports. The Chi-Square test indicated that there was no significant difference in the number of accidents among the different certification categories when either Controlled Flight into Terrain or Structural Failure was listed as cause. However, there was a significant difference in the frequency of accidents with regard to Loss of Control and Engine Failure accidents. The results of the ANCOVA test indicated that there was no significant difference in the accident rate with regard to Loss of Control, Controlled Flight into Terrain, or Structural Failure accidents. There was, however, a significant difference in Engine Failure accidents between Experimental-Amateur Built and the other categories.The text mining analysis of the narrative causes of Loss of Control accidents indicated that only the Civil Air Regulations 3 category airplanes had clusters of words associated with visual flight into instrument meteorological conditions. Civil Air Regulations 3 airplanes were designed and manufactured prior to the 1960s and in most cases have not been retrofitted to take advantage of newer technologies that could help prevent Loss of Control accidents. The study indicated that General Aviation aircraft certification rules do not have a statistically significant effect on aircraft accidents except for Loss of Control and Engine Failure. According to the literature, government oversight could have become an obstacle in the implementation of safety enhancing equipment that could reduce Loss of Control accidents. Oversight should focus on ensuring that Experimental-Amateur Built aircraft owners perform a functional test that could prevent some of the Engine Failure accidents.

ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

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

Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.

2002-07-01

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enablemore » much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)« less

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

Multispectral pyrometry for surface temperature measurement of oxidized Zircaloy claddings

NASA Astrophysics Data System (ADS)

Bouvry, B.; Cheymol, G.; Ramiandrisoa, L.; Javaudin, B.; Gallou, C.; Maskrot, H.; Horny, N.; Duvaut, T.; Destouches, C.; Ferry, L.; Gonnier, C.

2017-06-01

Non-contact temperature measurement in a nuclear reactor is still a huge challenge because of the numerous constraints to consider, such as the high temperature, the steam atmosphere, and irradiation. A device is currently developed at CEA to study the nuclear fuel claddings behavior during a Loss-of-Coolant Accident. As a first step of development, we designed and tested an optical pyrometry procedure to measure the surface temperature of nuclear fuel claddings without any contact, under air, in the temperature range 700-850 °C. The temperature of Zircaloy-4 cladding samples was retrieved at various temperature levels. We used Multispectral Radiation Thermometry with the hypothesis of a constant emissivity profile in the spectral ranges 1-1.3 μm and 1.45-1.6 μm. To allow for comparisons, a reference temperature was provided by a thermocouple welded on the cladding surface. Because of thermal losses induced by the presence of the thermocouple, a heat transfer simulation was also performed to estimate the bias. We found a good agreement between the pyrometry measurement and the temperature reference, validating the constant emissivity profile hypothesis used in the MRT estimation. The expanded measurement uncertainty (k = 2) of the temperature obtained by the pyrometry method was ±4 °C, for temperatures between 700 and 850 °C. Emissivity values, between 0.86 and 0.91 were obtained.

Thermal analyses of power subsystem components

NASA Technical Reports Server (NTRS)

Morehouse, Jeffrey H.

1990-01-01

The hiatus in the Space Shuttle (Orbiter) program provided time for an in-depth examination of all the subsystems and their past performance. Specifically, problems with reliability and/or operating limits were and continue to be of major engineering concern. The Orbiter Auxiliary Power Unit (APU) currently operates with electric resistance line heaters which are controlled with thermostats. A design option simplification of this heater subsystem is being considered which would use self-regulating heaters. A determination of the properties and thermal operating characteristics of these self-regulating heaters was needed. The Orbiter fuel cells are cooled with a freon loop. During a loss of external heat exchanger coolant flow, the single pump circulating the freon is to be left running. It was unknown what temperature and flow rate transient conditions of the freon would provide the required fuel cell cooling and for how long. The overall objective was the development of the thermal characterization and subsequent analysis of both the proposed self-regulating APU heater and the fuel cell coolant loop subsystem. The specific objective of the APU subsystem effort was to determine the feasibility of replacing the current heater and thermostat arrangement with a self-regulating heater. The specific objective of the fuel cell coolant subsystem work was to determine the tranient coolant temperature and associated flow rates during a loss-of-external heat exchanger flow.

A numerical study of the temperature field in a cooled radial turbine rotor

NASA Technical Reports Server (NTRS)

Hamed, A.; Baskharone, E.; Tabakoff, W.

1977-01-01

The three dimensional temperature distribution in the cooled rotor of a radial inflow turbine is determined numerically using the finite element method. Through this approach, the complicated geometries of the hot rotor and coolant passage surfaces are handled easily, and the temperatures are determined without loss of accuracy at these convective boundaries. Different cooling techniques with given coolant to primary flow ratios are investigated, and the corresponding rotor temperature fields are presented for comparison.

49 CFR 382.303 - Post-accident testing.

Code of Federal Regulations, 2013 CFR

2013-10-01

... functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who receives a citation within 8 hours of the occurrence under State or local law for a moving traffic violation... performing safety-sensitive functions with respect to the vehicle, if the accident involved the loss of human...

49 CFR 382.303 - Post-accident testing.

Code of Federal Regulations, 2010 CFR

2010-10-01

... functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who receives... performing safety-sensitive functions with respect to the vehicle, if the accident involved the loss of human life; or (2) Who receives a citation within thirty-two hours of the occurrence under State or local law...

Validation and Verification of Future Integrated Safety-Critical Systems Operating under Off-Nominal Conditions

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.

2010-01-01

Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents and reducing them will require a holistic integrated intervention capability. Future onboard integrated system technologies developed for preventing loss of vehicle control accidents must be able to assure safe operation under the associated off-nominal conditions. The transition of these technologies into the commercial fleet will require their extensive validation and verification (V and V) and ultimate certification. The V and V of complex integrated systems poses major nontrivial technical challenges particularly for safety-critical operation under highly off-nominal conditions associated with aircraft loss-of-control events. This paper summarizes the V and V problem and presents a proposed process that could be applied to complex integrated safety-critical systems developed for preventing aircraft loss-of-control accidents. A summary of recent research accomplishments in this effort is also provided.

Oil Spills

MedlinePlus

Oil spills often happen because of accidents, when people make mistakes or equipment breaks down. Other causes include natural disasters or deliberate acts. Oil spills have major environmental and economic effects. Oil ...

Curve Estimation of Number of People Killed in Traffic Accidents in Turkey

NASA Astrophysics Data System (ADS)

Berkhan Akalin, Kadir; Karacasu, Murat; Altin, Arzu Yavuz; Ergül, Bariş

2016-10-01

One or more than one vehicle in motion on the highway involving death, injury and loss events which have resulted are called accidents. As a result of increasing population and traffic density, traffic accidents continue to increase and this leads to both human losses and harm to the economy. In addition, also leads to social problems. As a result of increasing population and traffic density, traffic accidents continue to increase and this leads to both human losses and harm to the economy. In addition to this, it also leads to social problems. As a result of traffic accidents, millions of people die year by year. A great majority of these accidents occur in developing countries. One of the most important tasks of transportation engineers is to reduce traffic accidents by creating a specific system. For that reason, statistical information about traffic accidents which occur in the past years should be organized by versed people. Factors affecting the traffic accidents are analyzed in various ways. In this study, modelling the number of people killed in traffic accidents in Turkey is determined. The dead people were modelled using curve fitting method with the number of people killed in traffic accidents in Turkey dataset between 1990 and 2014. It was also predicted the number of dead people by using various models for the future. It is decided that linear model is suitable for the estimates.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

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

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

A liquid cooled garment temperature controller based on sweat rate

NASA Technical Reports Server (NTRS)

Chambers, A. B.; Blackaby, J. R.

1972-01-01

An automatic controller for liquid cooled space suits is reported that utilizes human sweat rate as the primary input signal. The controller is so designed that the coolant inlet temperature is inversely proportional to the subject's latent heat loss as evidenced by evaporative water loss.

Modeling of Kerena Emergency Condenser

NASA Astrophysics Data System (ADS)

Bryk, Rafał; Schmidt, Holger; Mull, Thomas; Wagner, Thomas; Ganzmann, Ingo; Herbst, Oliver

2017-12-01

KERENA is an innovative boiling water reactor concept equipped with several passive safety systems. For the experimental verification of performance of the systems and for codes validation, the Integral Test Stand Karlstein (INKA) was built in Karlstein, Germany. The emergency condenser (EC) system transfers heat from the reactor pressure vessel (RPV) to the core flooding pool in case of water level decrease in the RPV. EC is composed of a large number of slightly inclined tubes. During accident conditions, steam enters into the tubes and condenses due to the contact of the tubes with cold water at the secondary side. The condensed water flows then back to the RPV due to gravity. In this paper two approaches for modeling of condensation in slightly inclined tubes are compared and verified against experiments. The first approach is based on the flow regime map. Depending on the regime, heat transfer coefficient is calculated according to specific semi-empirical correlation. The second approach uses a general, fully-empirical correlation. The models are developed with utilization of the object-oriented Modelica language and the open-source OpenModelica environment. The results are compared with data obtained during a large scale integral test, simulating loss of coolant accident performed at Integral Test Stand Karlstein (INKA). The comparison shows a good agreement.Due to the modularity of models, both of them may be used in the future in systems incorporating condensation in horizontal or slightly inclined tubes. Depending on his preferences, the modeller may choose one-equation based approach or more sophisticated model composed of several exchangeable semi-empirical correlations.

Heat up and potential failure of BWR upper internals during a severe accident

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

Robb, Kevin R

2015-01-01

In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, andmore » relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less

Design of load-to-failure tests of high-voltage insulation breaks for ITER's cryogenic network

NASA Astrophysics Data System (ADS)

Langeslag, S. A. E.; Rodriguez Castro, E.; Aviles Santillana, I.; Sgobba, S.; Foussat, A.

2015-12-01

The development of new generation superconducting magnets for fusion research, such as the ITER experiment, is largely based on coils wound with so-called cable-in-conduit conductors. The concept of the cable-in-conduit conductor is based on a direct cooling principle, by supercritical helium, flowing through the central region of the conductor, in close contact with the superconducting strands. Consequently, a direct connection exists between the electrically grounded helium coolant supply line and the highly energised magnet windings. Various insulated regions, constructed out of high-voltage insulation breaks, are put in place to isolate sectors with different electrical potential. In addition to high voltages and significant internal helium pressure, the insulation breaks will experience various mechanical forces resulting from differential thermal contraction phenomena and electro-magnetic loads. Special test equipment was designed, prepared and employed to assess the mechanical reliability of the insulation breaks. A binary test setup is proposed, where mechanical failure is assumed when leak rate of gaseous helium exceeds 10-9·Pa·m3/s. The test consists of a load-to-failure insulation break charging, in tension, while immersed in liquid nitrogen at the temperature of 77 K. Leak tightness during the test is monitored by measuring the leak rate of the gaseous helium, directly surrounding the insulation break, with respect to the existing vacuum inside the insulation break. The experimental setup is proven effective, and various insulation breaks performed beyond expectations.

Design and testing of a self-actuated shut down system for the protection of liquid metal fast breeder reactors (LMFBRs)

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

Josephson, J.; Sowa, E.S.

1977-04-01

The design and testing of a simple and reliable Self-Actuated Shutdown System (SASS) for the protection of Liquid Metal Fast Breeder Reactors (LMFBRs) is described. A ferromagnetic Curie temperature permanent magnet holding device has been selected for the design of the Self-Actuated Shutdown System in order to enhance the safety of liquid metal cooled fast reactors (LMFBRs). The self-actuated, self-contained device operates such that accident conditions, resulting in increased coolant temperature or neutron flux reduce the magnetic holding force suspending a neutron absorber above the core by raising the temperature of the trigger mechanism above the Curie point. Neutron absorbermore » material is then inserted into the core, under gravity, terminating the accident. Two possible design variations of the selected concept are presented.« less

Nuclear reactor shutdown system

DOEpatents

Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

1981-01-01

An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

School Security and Safety.

ERIC Educational Resources Information Center

Licht, Kenneth F.

The author contends that safety and accident prevention should be given primary consideration in a school system's risk management program. He argues that accidents and losses are symptoms of defects in the management system. Two classes of loss discussed are (1) accidental -- injury/loss resulting from unintended events; and (2) purposeful --…

[Analisys of work-related accidents and incidents in an oil refinery in Rio de Janeiro].

PubMed

de Souza, Carlos Augusto Vaz; de Freitas, Carlos Machado

2003-01-01

Accidents in the chemical industry can have serious consequences for workers, communities, and the environment and are thus highly relevant to public health. This article is the result of an occupational surveillance project involving several public institutions. We analyze 800 work-related accidents that resulted in injuries, environmental damage, or loss of production in 1997 in an oil refinery located in Rio de Janeiro, Brazil. The methodology was based on managerial and organizational approaches to accident investigation, with the European Union reporting system as the reference. The results highlight various limitations in the process of reporting and investigating accidents, as well as a certain hierarchy of accidents, with more attention given to accidents involving loss of production and less to those resulting in injuries, particularly among outsourced workers.

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

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

Munganahalli, D.

Sedco Forex is a drilling contractor that operates approximately 80 rigs on land and offshore worldwide. The HSE management system developed by Sedco Forex is an effort to prevent accidents and minimize losses. An integral part of the HSE management system is establishing risk profiles and thereby minimizing risk and reducing loss exposures. Risk profiles are established based on accident reports, potential accident reports and other risk identification reports (RIR) like the Du Pont STOP system. A rig could fill in as many as 30 accident reports, 30 potential accident reports and 500 STOP cards each year. Statistics are importantmore » for an HSE management system, since they are indicators of success or failure of HSE systems. It is however difficult to establish risk profiles based on statistical information, unless tools are available at the rig site to aid with the analysis. Risk profiles are then used to identify important areas in the operation that may require specific attention to minimize the loss exposure. Programs to address the loss exposure can then be identified and implemented with either a local or corporate approach. In January 1995, Sedco Forex implemented a uniform HSE Database on all the rigs worldwide. In one year companywide, the HSE database would contain information on approximately 500 accident and potential accident reports, and 10,000 STOP cards. This paper demonstrates the salient features of the database and describes how it has helped in establishing key risk profiles. It also shows a recent example of how risk profiles have been established at the corporate level and used to identify the key contributing factors to hands and finger injuries. Based on this information, a campaign was launched to minimize the frequency of occurrence and associated loss attributed to hands and fingers accidents.« less

Structural response of transport airplanes in crash situations

NASA Technical Reports Server (NTRS)

Thomson, R. G.; Caiafa, C.

1983-01-01

This report highlights the results of contractural studies of transport accident data undertaken in a joint research program sponsored by the FAA and NASA. From these accident data studies it was concluded that the greatest potential for improved transport crashworthiness is in the reduction of fire related fatalities. Accident data pertaining to fuselage integrity, main landing gear collapse, fuel tank rupture, wing breaks, tearing of tank lower surfaces, and engine pod scrubbing are discussed. In those accidents where the energy absorbing protective capability of the fuselage structure is expended and the airplane experiences major structural damage, trauma caused fatalities are also discussed. The dynamic performance of current seat/restraint systems are examined but it is concluded that the accident data does not adequately define the relationship between occupant response and the dynamic interaction with the seat, floor and fuselage structure.

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

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

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

Development of Innovative Accident Tolerant High Thermal Conductivity UO 2-Diamond Composite Fuel Pellets

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

Tulenko, James; Subhash, Ghatu

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO 2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO 2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO 2 – SiC and UO 2 – Carbon Nanotube fuel pins. UF ismore » proving with the current research results that the addition of diamond micro particles to UO 2 may greatly enhanced the thermal conductivity of the UO 2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.« less

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.

[Accidents and violence in childhood and adolescence: risk and protective factors].

PubMed

Martins, Christine Baccarat de Godoy

2013-01-01

Singled out by statistics as the third leading cause of mortality in our country, external causes (accidents and violence) entail a great impact with economic, social and emotional rebound. Knowing the factors related to the event is essential, because it allows identifying and breaking the web that determines morbidity and mortality from external causes. The study aims to analyze the existing publications on the factors associated with accidents and violence, in order to provide theoretical support for professionals in their practices. This is a bibliographical study of the Liliacs, Medline and Scielo databanks. The knowledge of the risk and protection factors discussed in the present study enables subsidize the practice of social actors engaged in transforming the conditions that lead to accidents and violence.

Analysis of loss of decay-heat-removal sequences at Browns Ferry Unit One

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

Harrington, R.M.

1983-01-01

This paper summarizes the Oak Ridge National Laboratory (ORNL) report Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA). The purpose of the SASA studies is to predetermine the probable course of postulated severe accidents so as to establish the timing andmore » the sequence of events. The SASA studies also produce recommendations concerning the implementation of better system design and better emergency operating instructions and operator training. The ORNL studies also include a detailed, best-estimate calculation of the release and transport of radioactive fission products following postulated severe accidents.« less

Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

NASA Astrophysics Data System (ADS)

Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

1988-09-01

This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

Is "Summer" Reading Loss Universal? Using Ongoing Literacy Assessment in Malawi to Estimate the Loss from Grade-Transition Breaks

ERIC Educational Resources Information Center

Slade, Timothy S.; Piper, Benjamin; Kaunda, Zikani; King, Simon; Ibrahim, Hibatalla

2017-01-01

Summer learning loss--decreased academic performance following an extended school break, typically during the period after one grade ends and before another grade starts--is a well-documented phenomenon in North America, but poorly described in sub-Saharan African contexts. In this article, we use the term "grade-transition break" loss…

Effects of rotation on coolant passage heat transfer. Volume 2: Coolant passages with trips normal and skewed to the flow

NASA Technical Reports Server (NTRS)

Johnson, B. V.; Wagner, J. H.; Steuber, G. D.

1993-01-01

An experimental program was conducted to investigate heat transfer and pressure loss characteristics of rotating multipass passages, for configurations and dimensions typical of modem turbine blades. This experimental program is one part of the NASA Hot Section Technology (HOST) Initiative, which has as its overall objective the development and verification of improved analysis methods that will form the basis for a design system that will produce turbine components with improved durability. The objective of this program was the generation of a data base of heat transfer and pressure loss data required to develop heat transfer correlations and to assess computational fluid dynamic techniques for rotating coolant passages. The experimental work was broken down into two phases. Phase 1 consists of experiments conducted in a smooth wall large scale heat transfer model. A detailed discussion of these results was presented in volume 1 of a NASA Report. In Phase 2 the large scale model was modified to investigate the effects of skewed and normal passage turbulators. The results of Phase 2 along with comparison to Phase 1 is the subject of this Volume 2 NASA Report.

Modelling of LOCA Tests with the BISON Fuel Performance Code

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

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

NASA Astrophysics Data System (ADS)

Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

2014-06-01

In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

Sensory impairment and driving: the Blue Mountains Eye Study.

PubMed Central

Ivers, R Q; Mitchell, P; Cumming, R G

1999-01-01

OBJECTIVES: This study examined the associations between vision, hearing, loss, and car accidents. METHODS: A cross-sectional survey of 3654 people aged 49 years and older in the Blue Mountains, Australia, was used. Each subject had a detailed eye examination and interview. RESULTS: Self-reported car accident rates in the past year among 2379 current drivers were 5.6% for those aged 49 to 79 years and 9.1% for those 80 years and older. A 2-line difference in visual acuity was associated with increased risk of accidents (adjusted prevalence ratio [PR] = 1.6), as was visual acuity worse than 6/18 in the right eye (PR = 2.0), overall moderate hearing loss (PR = 1.9), and hearing loss in the right ear (PR = 1.8). CONCLUSIONS: Sensory loss in drivers may be an important risk factor for car accidents. PMID:9987472

Severe Accident Test Station Activity Report

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

Pint, Bruce A.; Terrani, Kurt A.

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accidentmore » Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.« less

[Insurance system. Prevention from viewpoint of the insurer].

PubMed

Brechtbühl, P

1978-12-01

The purpose of an insurance must not be restricted to the payment of claims to those insured persons who suffered a loss, for loss prevention is much preferable to claim settlement. A whole range of different institutions and measures has been established by the Swiss insurers, in which many insurance branches participate. The loss preventing activities can be listed as follows:--Activities of the fire insurers to prevent and fight fires. This is the prevailing duty of the Consulting Agency for Fire Prevention (BfB) as well as the Fire Prevention Service for Industry and Trade (BVD).--Activities of the accident insurers to prevent accidents. The fight against accidents, mostly traffic accidents, in sports and at home is the foremost task of the Swiss Council for the Prevention of Accidents (BfU), an institution created by the Conference of Accident Insurance Managers (UDK) and the Swiss National Accident Insurance Fund (SUVA).--The Health Service in life insurance, after all the periodical medical examinations and consultations granted by many insurers to their insured persons, as well as the pamphlets aiming at health education published by several Companies and finally institutions and measures to promote fitness, e.g. VITA-Parcours.

Supercritical Brayton Cycle Nuclear Power System Concepts

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6H14, Tcritical = 506.1 K) provided they have adequate chemical compatibility and stability. Overall the use of supercritical Brayton cycles may offer ``break through'' operating capabilities for space nuclear power plants because high efficiencies can be achieved a very low reactor operating temperatures which in turn allows for the use of available fuels, cladding, and structural materials.

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

Roussel, G.

Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbersmore » at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.« less

Heat up and failure of BWR upper internals during a severe accident

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

Robb, Kevin R.

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Heat up and failure of BWR upper internals during a severe accident

DOE PAGES

Robb, Kevin R.

2017-02-21

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplifiedmore » heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.« less

Experimental investigation of temperature rise in bone drilling with cooling: A comparison between modes of without cooling, internal gas cooling, and external liquid cooling.

PubMed

Shakouri, Ehsan; Haghighi Hassanalideh, Hossein; Gholampour, Seifollah

2018-01-01

Bone fracture occurs due to accident, aging, and disease. For the treatment of bone fractures, it is essential that the bones are kept fixed in the right place. In complex fractures, internal fixation or external methods are used to fix the fracture position. In order to immobilize the fracture position and connect the holder equipment to it, bone drilling is required. During the drilling of the bone, the required forces to chip formation could cause an increase in the temperature. If the resulting temperature increases to 47 °C, it causes thermal necrosis of the bone. Thermal necrosis decreases bone strength in the hole and, subsequently, due to incomplete immobilization of bone, fracture repair is not performed correctly. In this study, attempts have been made to compare local temperature increases in different processes of bone drilling. This comparison has been done between drilling without cooling, drilling with gas cooling, and liquid cooling on bovine femur. Drilling tests with gas coolant using direct injection of CO 2 and N 2 gases were carried out by internal coolant drill bit. The results showed that with the use of gas coolant, the elevation of temperature has limited to 6 °C and the thermal necrosis is prevented. Maximum temperature rise reached in drilling without cooling was 56 °C, using gas and liquid coolant, a maximum temperature elevation of 43 °C and 42 °C have been obtained, respectively. This resulted in decreased possibility of thermal necrosis of bone in drilling with gas and liquid cooling. However, the results showed that the values obtained with the drilling method with direct gas cooling are independent of the rotational speed of drill.

Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

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

Szilard, Ronaldo H.; Youngblood, Robert W.; Zhang, Hongbin

2015-09-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions formore » NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.« less

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

NASA Medical Response to Human Spacecraft Accidents

NASA Technical Reports Server (NTRS)

Patlach, Robert

2010-01-01

Manned space flight is risky business. Accidents have occurred and may occur in the future. NASA's manned space flight programs, with all their successes, have had three fatal accidents, one at the launch pad and two in flight. The Apollo fire and the Challenger and Columbia accidents resulted in a loss of seventeen crewmembers. Russia's manned space flight programs have had three fatal accidents, one ground-based and two in flight. These accidents resulted in the loss of five crewmembers. Additionally, manned spacecraft have encountered numerous close calls with potential for disaster. The NASA Johnson Space Center Flight Safety Office has documented more than 70 spacecraft incidents, many of which could have become serious accidents. At the Johnson Space Center (JSC), medical contingency personnel are assigned to a Mishap Investigation Team. The team deploys to the accident site to gather and preserve evidence for the Accident Investigation Board. The JSC Medical Operations Branch has developed a flight surgeon accident response training class to capture the lessons learned from the Columbia accident. This presentation will address the NASA Mishap Investigation Team's medical objectives, planned response, and potential issues that could arise subsequent to a manned spacecraft accident. Educational Objectives are to understand the medical objectives and issues confronting the Mishap Investigation Team medical personnel subsequent to a human space flight accident.

26 CFR 1.846-3 - Fresh start and reserve strengthening.

Code of Federal Regulations, 2012 CFR

2012-04-01

... undiscounted unpaid losses of $800,000 in the automobile liability line of business for the 1983 accident year... 1987 accident year and prior accident years (see section 1023(e)(2) of the 1986 Act); and (ii) By applying those discount factors as if the 1986 accident year were the 1987 accident year. (2) Example. The...

26 CFR 1.846-3 - Fresh start and reserve strengthening.

Code of Federal Regulations, 2011 CFR

2011-04-01

... undiscounted unpaid losses of $800,000 in the automobile liability line of business for the 1983 accident year... 1987 accident year and prior accident years (see section 1023(e)(2) of the 1986 Act); and (ii) By applying those discount factors as if the 1986 accident year were the 1987 accident year. (2) Example. The...

26 CFR 1.846-3 - Fresh start and reserve strengthening.

Code of Federal Regulations, 2014 CFR

2014-04-01

... undiscounted unpaid losses of $800,000 in the automobile liability line of business for the 1983 accident year... 1987 accident year and prior accident years (see section 1023(e)(2) of the 1986 Act); and (ii) By applying those discount factors as if the 1986 accident year were the 1987 accident year. (2) Example. The...

26 CFR 1.846-3 - Fresh start and reserve strengthening.

Code of Federal Regulations, 2013 CFR

2013-04-01

... undiscounted unpaid losses of $800,000 in the automobile liability line of business for the 1983 accident year... 1987 accident year and prior accident years (see section 1023(e)(2) of the 1986 Act); and (ii) By applying those discount factors as if the 1986 accident year were the 1987 accident year. (2) Example. The...

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

Outline of the Fukushima Daiichi Accident. Lessons Learned and Safety Enhancements

NASA Astrophysics Data System (ADS)

Hirano, Masashi

2017-09-01

Abstract. On March 11, 2011, an earthquake and subsequent tsunamis off the Pacific coastline of Japan's Tohoku region caused widespread devastation in Japan. As of June 10, 2016, it is reported that a total of 15,894 people lost their lives and 2,558 people are still unaccounted for. In Fukushima Prefecture, approximately 100,000 people are still obliged to live away from their homes due to the earthquake and tsunami as well as the Fukushima Daiichi accident. On the day, the earthquake and tsunami caused severe damages to the Tokyo Electric Power Company (TEPCO)'s Fukushima Daiichi Nuclear Power Station (NPS). All the units in operation, namely Units 1 to 3, were automatically shut down on seismic reactor protection system trips but the earthquake led to the loss of all off-site electrical power supplies to that site. The subsequent tsunami inundated the site up to 4 to 5 m above its ground level and caused, in the end, the loss of core cooling function in Units 1 to 3, resulting in severe core damages and containment vessel failures in these three units. Hydrogen was released from the containment vessels, leading to explosions in the reactor buildings of Units 1, 3 and 4. Radioactive materials were released to the atmosphere and were deposited on the land and in the ocean. One of the most important lessons learned is an importance to prevent such large scale common cause failures due to extreme natural events. This leads to a conclusion that application of the defense-in-depth philosophy be enhanced because the defense-in-depth philosophy has been and continues to be an effective way to account for uncertainties associated with risks. From the human and organizational viewpoints, the final report from the Investigation Committee of the Government pointed out so-called "safety myth" that existed among nuclear operators including TEPCO as well as the government, that serious severe accidents could never occur in nuclear power plants in Japan. After the accident, the Nuclear Regulation Authority (NRA) was established on September 19, 2012. The NRA very urgently developed and issued the new regulatory requirements on July 8, 2014, taking into the account the lessons learned from the accident. It is noted that the NRA issued the Statement of Nuclear Safety Culture on May 27, 2015 which clearly expressed the NRA's commitment to break with the safety myth. This paper briefly presents the outline of the Fukushima Daiichi accident and summarizes the major lessons learned having been drawn and safety enhancements having been done in Japan for the purpose of giving inputs to the discussions to be taken place in the Special Invited Session "Fukushima, 5 years after".

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

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

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

Structures for handling high heat fluxes

NASA Astrophysics Data System (ADS)

Watson, R. D.

1990-12-01

The divertor is reconized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes > 5 MW/m 2. High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40-60 MW/m 2. Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1-10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components.

Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

NASA Astrophysics Data System (ADS)

Philipose, K.; Shenton, B.

2011-04-01

The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

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

Embitterment and bereavement: The Sewol ferry accident example.

PubMed

Chae, Jeong-Ho; Huh, Hyu Jung; Choi, Won Joon

2018-01-01

On Wednesday, April 16, 2014, 261 high school students on a field trip died in the sinking of the Sewol ferry. The bereaved family of the Sewol ferry accident experienced one of the most painful traumatic losses such as the sudden death of one's child through an accident. This article reviewed and discussed embitterment related to traumatic loss through the example of the Sewol ferry accident. Embitterment-related issues and problems in coping with the accident that is caused by societal factors were described. In addition, embitterment-related findings of several previous studies based on bereaved families' mental health cohort study were reviewed. Traumatic loss of the human-made ferry accident was accompanied with feelings of being cheated, injustice, incompetence, wrongdoing by a perpetrator, and the destruction of one's belief and value system, causing severe embitterment. Embitterment was related to other mental health problems including depression, anxiety, and complicated grief. Social support and positive individual resource including optimism and wisdom can be helpful for recovery from posttraumatic embitterment. The goal of grief is to remember the decedent, understand the changes created by the loss, and determine how to reinvest in life. Embitterment may disturb the process of grief. Without the management of the embitterment, true grief may not be possible. The breakdown of value systems and severe embitterment should get more attention in future research. (PsycINFO Database Record (c) 2018 APA, all rights reserved).

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.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

NASA Astrophysics Data System (ADS)

Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

2016-10-01

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

DOE PAGES

Terrani, K. A.; Pint, B. A.; Kim, Y. -J.; ...

2016-06-29

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation ofmore » very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.« less

Pilot simulation tests of propulsion control as backup to loss of primary flight controls for a mid-size jet transport

DOT National Transportation Integrated Search

1995-12-01

Partial failures of aircraft primary flight-control systems and structural : damages to aircraft during flight have led to catastrophic accidents with : subsequent loss of life. These accidents can be prevented if sufficient : alternate control autho...

Cost effective safety improvements for two-lane rural roads

DOT National Transportation Integrated Search

2008-02-01

Traffic accidents cause loss of life and property. Proper identification of accident causal factors is essential for composing countermeasures against traffic accidents and reducing related costs. However, two-lane rural roads have distinctive roadwa...

Vacuum-barrier window for wide-bandwidth high-power microwave transmission

DOEpatents

Caplan, M.; Shang, C.C.

1996-08-20

A vacuum output window comprises a planar dielectric material with identical systems of parallel ridges and valleys formed in opposite surfaces. The valleys in each surface neck together along parallel lines in the bulk of the dielectric. Liquid-coolant conduits are disposed linearly along such lines of necking and have water or even liquid nitrogen pumped through to remove heat. The dielectric material can be alumina, or its crystalline form, sapphire. The electric-field of a broadband incident megawatt millimeter-wave radio frequency energy is oriented perpendicular to the system of ridges and valleys. The ridges, about one wavelength tall and with a period of about one wavelength, focus the incident energy through in ribbons that squeeze between the liquid-coolant conduits without significant losses over very broad bands of the radio spectrum. In an alternative embodiment, the liquid-coolant conduits are encased in metal within the bulk of the dielectric. 4 figs.

Vacuum-barrier window for wide-bandwidth high-power microwave transmission

DOEpatents

Caplan, Malcolm; Shang, Clifford C.

1996-01-01

A vacuum output window comprises a planar dielectric material with identical systems of parallel ridges and valleys formed in opposite surfaces. The valleys in each surface neck together along parallel lines in the bulk of the dielectric. Liquid-coolant conduits are disposed linearly along such lines of necking and have water or even liquid nitrogen pumped through to remove heat. The dielectric material can be alumina, or its crystalline form, sapphire. The electric-field of a broadband incident megawatt millimeter-wave radio frequency energy is oriented perpendicular to the system of ridges and valleys. The ridges, about one wavelength tall and with a period of about one wavelength, focus the incident energy through in ribbons that squeeze between the liquid-coolant conduits without significant losses over very broad bands of the radio spectrum. In an alternative embodiment, the liquid-coolant conduits are encased in metal within the bulk of the dielectric.

A Comprehensive Analysis of the X-15 Flight 3-65 Accident

NASA Technical Reports Server (NTRS)

Dennehy, Cornelius J.; Orr, Jeb S.; Barshi, Immanuel; Statler, Irving C.

2014-01-01

The November 15, 1967, loss of X-15 Flight 3-65-97 (hereafter referred to as Flight 3-65) was a unique incident in that it was the first and only aerospace flight accident involving loss of crew on a vehicle with an adaptive flight control system (AFCS). In addition, Flight 3-65 remains the only incidence of a single-pilot departure from controlled flight of a manned entry vehicle in a hypersonic flight regime. To mitigate risk to emerging aerospace systems, the NASA Engineering and Safety Center (NESC) proposed a comprehensive review of this accident. The goal of the assessment was to resolve lingering questions regarding the failure modes of the aircraft systems (including the AFCS) and thoroughly analyze the interactions among the human agents and autonomous systems that contributed to the loss of the pilot and aircraft. This document contains the outcome of the accident review.

Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

NASA Astrophysics Data System (ADS)

Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

2015-10-01

In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

Noise and neurotoxic chemical exposure relationship to workplace traumatic injuries: A review☆

PubMed Central

Estill, Cheryl Fairfield; Rice, Carol H.; Morata, Thais; Bhattacharya, Amit

2017-01-01

Introduction More than 5,000 fatalities and eight million injuries occurred in the workplace in 2007 at a cost of $6 billion and $186 billion, respectively. Neurotoxic chemicals are known to affect central nervous system functions among workers, which include balance and hearing disorders. However, it is not known if there is an association between exposure to noise and solvents and acute injuries. Method A thorough review was conducted of the literature on the relationship between noise or solvent exposures and hearing loss with various health outcomes. Results The search resulted in 41 studies. Health outcomes included: hearing loss, workplace injuries, absence from work due to sickness, fatalities, hospital admissions due to workplace accidents, traffic accidents, hypertension, balance, slip, trips, or falls, cognitive measures, or disability retirement. Important covariates in these studies were age of employee, type of industry or occupation, or length of employment. Discussion Most authors that evaluated noise exposure concluded that higher exposure to noise resulted in more of the chosen health effect but the relationship is not well understood. Studies that evaluated hearing loss found that hearing loss was related to occupational injury, disability retirement, or traffic accidents. Studies that assessed both noise exposure and hearing loss as risk factors for occupational injuries reported that hearing loss was related to occupational injuries as much or more than noise exposure. Evidence suggests that solvent exposure is likely to be related to accidents or other health consequences such balance disorders. Conclusions Many authors reported that noise exposures and hearing loss, respectively, are likely to be related to occupational accidents. Practical applications The potential significance of the study is that findings could be used by managers to reduce injuries and the costs associated with those injures. PMID:28160812

Noise and neurotoxic chemical exposure relationship to workplace traumatic injuries: A review.

PubMed

Estill, Cheryl Fairfield; Rice, Carol H; Morata, Thais; Bhattacharya, Amit

2017-02-01

More than 5,000 fatalities and eight million injuries occurred in the workplace in 2007 at a cost of $6 billion and $186 billion, respectively. Neurotoxic chemicals are known to affect central nervous system functions among workers, which include balance and hearing disorders. However, it is not known if there is an association between exposure to noise and solvents and acute injuries. A thorough review was conducted of the literature on the relationship between noise or solvent exposures and hearing loss with various health outcomes. The search resulted in 41 studies. Health outcomes included: hearing loss, workplace injuries, absence from work due to sickness, fatalities, hospital admissions due to workplace accidents, traffic accidents, hypertension, balance, slip, trips, or falls, cognitive measures, or disability retirement. Important covariates in these studies were age of employee, type of industry or occupation, or length of employment. Most authors that evaluated noise exposure concluded that higher exposure to noise resulted in more of the chosen health effect but the relationship is not well understood. Studies that evaluated hearing loss found that hearing loss was related to occupational injury, disability retirement, or traffic accidents. Studies that assessed both noise exposure and hearing loss as risk factors for occupational injuries reported that hearing loss was related to occupational injuries as much or more than noise exposure. Evidence suggests that solvent exposure is likely to be related to accidents or other health consequences such balance disorders. Many authors reported that noise exposures and hearing loss, respectively, are likely to be related to occupational accidents. Practical applications: The potential significance of the study is that findings could be used by managers to reduce injuries and the costs associated with those injures. Published by Elsevier Ltd.

Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

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

Wichner, R.P.; Weber, C.F.; Hodge, S.A.

1984-01-01

This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris ontomore » the drywell floor.« less

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

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

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun

2015-07-01

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to dependmore » largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.« less

Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

NASA Astrophysics Data System (ADS)

Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

2015-08-01

The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations under the effect of external periodic loads caused by an earthquake when the vibration frequency of the reactor plant main equipment and the frequency of elastic waves fall in the frequency band corresponding to the maximal values of envelope response spectra.

Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments

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

Parish, Chad M.; Terrani, Kurt A.; Kim, Young -Jin

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corrodedmore » region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA – which suffered the least corrosion –retained some Al in the corroded region. Lastly, the results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.« less

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

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

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

Steam Oxidation Testing in the Severe Accident Test Station

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

Pint, Bruce A.; McMurray, Jake W.

2016-08-01

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the coremore » during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.« less

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

Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HXmore » channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.« less

Technical challenges of upset recovery training : simulating the element of surprise

DOT National Transportation Integrated Search

2010-07-30

This invited paper is written in the context of a concerted effort by the aviation industry and regulators to reduce the occurrence of Loss of Control (LOC) accidents. LOC accidents have taken the lead among fatal airplane accidents, recently outpaci...

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.

[Mechanisms and prevention of windsurfing injuries].

PubMed

Petersen, W; Rau, J; Hansen, U; Zantop, T; Stein, V

2003-09-01

Aim of this study was to analyse the mechanisms of windsurfing injuries. For this purpose we performed a internet based survey among 327 windsurfers in Germany. Overall 630 accidents have been registered among all 327 athletes during the 2000 season. The majority of injuries were classified as minor injury. The most common injury was the bruise. 70 participants reported fractures, 26 participants ruptured a ligament. 280 injuries required medical treatment; in 67 cases even surgical treatment was necessary. The majority of accidents happened at wind power of 5-6 Beaufort after 2 hour exercise. A technical mistake was the most frequent cause for the accident. The most dangerous manoeuvres were difficult jumps (e. g. front loop, backward loop, 70 injuries). In 46 cases the weather conditions were underestimated. Only 10 windsurfers reported about broken material as cause for the injury. One half of the injuries happened in wave conditions. The analysis of injury mechanisms allows conclusions regarding injury prevention. A longer break after 60 minutes windsurfing might help to prevent injuries due to poor physical fitness. After one hour windsurfing without a break training of difficult manoeuvres should not be performed. The use of a helmet might prevent head injuries during training of difficult jumps. "Overpower situations" should be prevented by choosing the right board and sail size.

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

NASA Astrophysics Data System (ADS)

Gussev, M. N.; Cakmak, E.; Field, K. G.

2018-06-01

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

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

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance

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

Williford, R.E.

1986-09-01

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperaturesmore » above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.« less

Inspection of Nuclear Power Plant Containment Structures

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

Graves, H.L.; Naus, D.J.; Norris, W.E.

1998-12-01

Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discoveredmore » at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.« 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

The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

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

Repetto, G.; Dominguez, C.; Durville, B.

The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the balloonedmore » areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.« less

TH-CD-BRA-09: Towards Absolute Dose Measurement in MRI-Linac and Gamma-Knife: Design and Construction of An MR-Compatible Water Calorimeter

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

Entezari, N; Sarfehnia, A; Renaud, J

Purpose: The purpose of this work is to design and optimize a portable Water Calorimeter (WC) for use in a commercial MRI-linac and Gamma-knife in addition to conventional radiotherapy linacs. Water calorimeters determine absorbed dose to water at a point by measuring radiation-induced temperature rise of the volume (the two are related by the medium specific heat capacity). In this formalism, one important correction factor is heat transfer correction k-ht. It compensates for heat gain/loss due to conductive and convective effects, and is numerically calculated as ratio of temperature rise in the absence of heat loss to that in themore » presence of heat loss. Operating at 4°C ensures convection is minimal. Methods: A commercial finite element software was used to evaluate several WC designs with different insulation materials and thicknesses; channels allowing coolant to travel around WC (to sustain WC at 4°C) were modeled, and worst-case scenario variation in the temperature of the coolant was simulated for optimization purposes (2.6 mK/s). Additionally, several calorimeter vessel design parameters (front/back glass thickness/separation, diameter) were also simulated and optimized. Optimization is based on minimizing long term calorimeter drift (24h) as well as variation and magnitude of k-ht. Results: The final selected WC design reached a modest drift of 11µK/s after 15h for the worst-case coolant temperature variation. This design consists of coolant channels being encompassed on both sides by cryogel insulation. For the MRI-linac beam, glass thickness plays the largest effect on k-ht with variation of upto 0.6% in the first run for thicknesses ranging between 0.5–1.7mm. Subsequent runs vary only within 0.1% with glass thickness. Other factors such as vessel radius and top/bottom glass separation have sub 0.1% effects on k-ht. Conclusion: An MR-safe 4°C stagnant WC appropriate for dosimetry in MRI-linac and Gamma-Knife was designed, optimized, and construction is nearly completed. NSERC Discovery Grant RGPIN-435608.« less

Monte Carlo simulation of single accident airport risk profile

NASA Technical Reports Server (NTRS)

1979-01-01

A computer simulation model was developed for estimating the potential economic impacts of a carbon fiber release upon facilities within an 80 kilometer radius of a major airport. The model simulated the possible range of release conditions and the resulting dispersion of the carbon fibers. Each iteration of the model generated a specific release scenario, which would cause a specific amount of dollar loss to the surrounding community. By repeated iterations, a risk profile was generated, showing the probability distribution of losses from one accident. Using accident probability estimates, the risks profile for annual losses was derived. The mechanics are described of the simulation model, the required input data, and the risk profiles generated for the 26 large hub airports.

100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Mason, Lee S.

1992-01-01

Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

Sleep loss and accidents--work hours, life style, and sleep pathology.

PubMed

Akerstedt, Torbjörn; Philip, Pierre; Capelli, Aurore; Kecklund, Göran

2011-01-01

A very important outcome of reduced sleep is accidents. The present chapter will attempt to bring together some of the present knowledge in this area. We will focus on the driving situation, for which the evidence of the link between sleep loss and accidents is quite well established, but we will also bring up working life in general where evidence is more sparse. It should be emphasized that reduced sleep as a cause of accidents implies that the mediating factor is sleepiness (or fatigue). This link is discussed elsewhere in this volume, but here we will bring in sleepiness (subjective or physiological) as an explanatory factor of accidents. Another central observation is that many real life accident studies do not link accidents to reduced sleep, but infer reduced sleep and/or sleepiness from the context, like, for example, from work schedules, life styles, or sleep pathology. Reduced sleep is mainly due to suboptimal work schedules (or to a suboptimal life style) or to sleep pathology. We have divided the present chapter into two areas. Copyright © 2011 Elsevier B.V. All rights reserved.

Summary of reactor plant conditions during L2-2 pre-LOCE maneuver

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

Tsang, F.Y.; Yarbrough, W.M.; Cannon, J.W.

1979-04-26

This document presents the experimental results obtained during the pre-Loss of Coolant Experiment (LOCE) manuever and the core conditions prior to the L2-2 LOCE. The peak linear heat rate prior to the blowdown was 8.04 kW/ft, the primary coolant mass flow rate was 1.539 Mlbm/hr, the hot leg temperature was 585.9/sup 0/F, and the core ..delta..T was 42/sup 0/F. These conditions satisfied the requirements specified for the L2-2 LOCE except for the hot leg temperature being 12/sup 0/F below the desired 598/sup 0/F.

Life events and borderline personality features: the influence of gene-environment interaction and gene-environment correlation.

PubMed

Distel, M A; Middeldorp, C M; Trull, T J; Derom, C A; Willemsen, G; Boomsma, D I

2011-04-01

Traumatic life events are generally more common in patients with borderline personality disorder (BPD) than in non-patients or patients with other personality disorders. This study investigates whether exposure to life events moderates the genetic architecture of BPD features. As the presence of genotype-environment correlation (rGE) can lead to spurious findings of genotype-environment interaction (G × E), we also test whether BPD features increase the likelihood of exposure to life events. The extent to which an individual is at risk to develop BPD was assessed with the Personality Assessment Inventory - Borderline features scale (PAI-BOR). Life events under study were a divorce/break-up, traffic accident, violent assault, sexual assault, robbery and job loss. Data were available for 5083 twins and 1285 non-twin siblings. Gene-environment interaction and correlation were assessed by using structural equation modelling (SEM) and the co-twin control design. There was evidence for both gene-environment interaction and correlation. Additive genetic influences on BPD features interacted with the exposure to sexual assault, with genetic variance being lower in exposed individuals. In individuals who had experienced a divorce/break-up, violent assault, sexual assault or job loss, environmental variance for BPD features was higher, leading to a lower heritability of BPD features in exposed individuals. Gene-environment correlation was present for some life events. The genes that influence BPD features thus also increased the likelihood of being exposed to certain life events. To our knowledge, this study is the first to test the joint effect of genetic and environmental influences and the exposure to life events on BPD features in the general population. Our results indicate the importance of both genetic vulnerability and life events.

Posttest analysis of LOFT LOCE L2-3 using the ESA RELAP4 blowdown model. [PWR

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

Perryman, J.L.; Samuels, T.K.; Cooper, C.H.

A posttest analysis of the blowdown portion of Loss-of-Coolant Experiment (LOCE) L2-3, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed using the experiment safety analysis (ESA) RELAP4/MOD5 computer model. Measured experimental parameters were compared with the calculations in order to assess the conservatisms in the ESA RELAP4/MOD5 model.

Epidemiological patterns of ocular trauma.

PubMed

Thylefors, B

1992-05-01

Ocular trauma is the cause of blindness in approximately half a million people worldwide, and many more have suffered partial loss of sight. Trauma is often the most important cause of unilateral loss of vision, particularly in developing countries. There is a cumulative risk of ocular trauma and visual loss during life, but the true incidence of accidents involving the eyes is not known. Males tend to have more eye trauma than females, and this is already apparent from childhood; lower socioeconomic classes are also more associated with ocular trauma. The setting for the occurrence of trauma is most commonly the workplace and, increasingly, road accidents. On the other hand, domestic accidents are probably under-reported. Of particular importance in some developing countries is the occurrence of superficial corneal trauma in agricultural work, often leading to rapidly progressing corneal ulceration and visual loss. The impact of ocular trauma, in terms of need for medical care, loss of income and cost of rehabilitation services when indicated, clearly makes the strengthening of preventive measures very worthwhile.

Steam Oxidation Testing in the Severe Accident Test Station

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

Pint, Bruce A.

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative wouldmore » significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.« less

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

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

Joy L. Rempe; Darrell L. Knudson

2013-03-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensorsmore » that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.« less

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

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

Joy L. Rempe; Darrell L. Knudson

2014-05-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensorsmore » that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.« less

Human and fishing vessel losses in sea accidents in the UK fishing industry from 1948 to 2008.

PubMed

Roberts, Stephen E; Jaremin, Bogdan; Marlow, Peter B

2010-01-01

To investigate long-term trends in mortality rates for accidents to fishing vessels in the UK fishing industry from 1948 to 2008; to investigate the circumstances and causes of these fishing vessel accidents and trends in fishing vessel losses. Examination of paper death inquiry files, death registers, marine accident investigative files, annual casualty and death returns. Of 1039 fatalities from accidents to UK fishing vessels from 1948 to 2008, most (65%) resulted from vessels that foundered (or capsized or disappeared), followed by vessels grounding (21%), collisions (7%), and fires and explosions (5%). There was a significant increase over time of 1.04% per year in the overall fishing vessel loss rate and for vessels that foundered (5.19%), a reduction for vessels grounding (1.13%), but no trends for collisions or fires and explosions. Regarding mortality, there was a significant reduction over time for grounding (1.44%) and a non-significant reduction for vessel accidents overall, but no trends for other types of vessel accident. Mortality was highest during the winter months (for foundering and grounding), during night time (for grounding, fires and explosions), and afternoons (foundering and collisions). Since 1976, most fatalities from collisions (83%) occurred in the English Channel and North Sea, while 49% from grounding occurred off the west coast of Scotland. The mortality rate from fishing vessel casualties in UK fishing is still very high. Fatalities in recent years have often been linked to fishing vessels that are unstable, overloaded, and unseaworthy.

Epidemiology of school accidents during a six school-year period in one region in Poland.

PubMed

Sosnowska, Stefania; Kostka, Tomasz

2003-01-01

The aim of the study was to analyse the incidence of school accidents in relation to school size, urban/rural environment and conditions of physical education classes. 202 primary schools with nearly 50,000 students aged 7-15 years were studied during a 6-year period in the Włocławek region in Poland. There were in total 3274 school accidents per 293,000 student-years. Accidents during breaks (36.6%) and physical education (33.2%) were most common. Most frequently accidents took place at schoolyard (29.7%), gymnasium (20.2%), and in the corridor and stairs (25.2%). After adjustment for students' age and sex, student-staff ratio and duration of school hours, urban environment increased the probability of accident (OR: 1.25; 95% CI: 1.14-1.38). Middle-size schools (8-23 classes) had similar accident rate as small schools (OR: 0.93; 95% CI: 0.83-1.04), while schools with 24-32 classes (OR: 1.26; 95% CI: 1.10-1.43) and with > or = 33 classes (OR: 1.36; 95% CI: 1.17-1.58) had increased accident rate. Presence of a gymnasium was also associated with increased probability of accident (OR: 1.49; 95% CI: 1.38-1.61). Urban environment, larger school-size and equipment with full-size gymnasium are important and independent risk factors for school accidents. These findings provide some new insights into the epidemiology of school-related accidents and may be useful information for the planning of strategies to reduce accident incidence in schools.

Performance assessment of conventional and base-isolated nuclear power plants for earthquake and blast loadings

NASA Astrophysics Data System (ADS)

Huang, Yin-Nan

Nuclear power plants (NPPs) and spent nuclear fuel (SNF) are required by code and regulations to be designed for a family of extreme events, including very rare earthquake shaking, loss of coolant accidents, and tornado-borne missile impacts. Blast loading due to malevolent attack became a design consideration for NPPs and SNF after the terrorist attacks of September 11, 2001. The studies presented in this dissertation assess the performance of sample conventional and base isolated NPP reactor buildings subjected to seismic effects and blast loadings. The response of the sample reactor building to tornado-borne missile impacts and internal events (e.g., loss of coolant accidents) will not change if the building is base isolated and so these hazards were not considered. The sample NPP reactor building studied in this dissertation is composed of containment and internal structures with a total weight of approximately 75,000 tons. Four configurations of the reactor building are studied, including one conventional fixed-base reactor building and three base-isolated reactor buildings using Friction Pendulum(TM), lead rubber and low damping rubber bearings. The seismic assessment of the sample reactor building is performed using a new procedure proposed in this dissertation that builds on the methodology presented in the draft ATC-58 Guidelines and the widely used Zion method, which uses fragility curves defined in terms of ground-motion parameters for NPP seismic probabilistic risk assessment. The new procedure improves the Zion method by using fragility curves that are defined in terms of structural response parameters since damage and failure of NPP components are more closely tied to structural response parameters than to ground motion parameters. Alternate ground motion scaling methods are studied to help establish an optimal procedure for scaling ground motions for the purpose of seismic performance assessment. The proposed performance assessment procedure is used to evaluate the vulnerability of the conventional and base-isolated NPP reactor buildings. The seismic performance assessment confirms the utility of seismic isolation at reducing spectral demands on secondary systems. Procedures to reduce the construction cost of secondary systems in isolated reactor buildings are presented. A blast assessment of the sample reactor building is performed for an assumed threat of 2000 kg of TNT explosive detonated on the surface with a closest distance to the reactor building of 10 m. The air and ground shock waves produced by the design threat are generated and used for performance assessment. The air blast loading to the sample reactor building is computed using a Computational Fluid Dynamics code Air3D and the ground shock time series is generated using an attenuation model for soil/rock response. Response-history analysis of the sample conventional and base isolated reactor buildings to external blast loadings is performed using the hydrocode LS-DYNA. The spectral demands on the secondary systems in the isolated reactor building due to air blast loading are greater than those for the conventional reactor building but much smaller than those spectral demands associated with Safe Shutdown Earthquake shaking. The isolators are extremely effective at filtering out high acceleration, high frequency ground shock loading.

Experimental evaluation of a translating nozzle sidewall radial turbine

NASA Technical Reports Server (NTRS)

Roelke, Richard J.; Rogo, Casimir

1987-01-01

An experimental performance evaluation was made of two movable sidewall variable area radial turbines. The turbine designs were representative of the gas generator turbine of a variable flow capacity rotorcraft engine. The first turbine was an uncooled design while the second turbine had a cooled nozzle but an uncooled rotor. The cooled nozzle turbine was evaluated both with and without coolant flow. The test results showed that the movable nozzle wall is a viable and efficient means to effectively control the flow capacity of a radial turbine. Peak efficiencies of the second turbine with and without nozzle coolant were 86.5 and 88 percent respectively. These values are comparable to pivoting vane variable geometry turbines; however, the decrease in efficiency as the flow was varied from the design value was much less for the movable wall turbine. Several design improvements which should increase the turbine efficiency one or two more points are identified. These design improvements include reduced leakage losses and relocation of the vane coolant ejection holes to reduce mainstream disturbance.

76 FR 69320 - Agency Request for Reinstatement of a Previously Approved Information Collection(s): Aircraft...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-08

... Previously Approved Information Collection(s): Aircraft Accident Liability Insurance AGENCY: Office of the...: Aircraft Accident Liability Insurance. Form Numbers: OST Forms 6410 and 6411. Type of Review: Reinstatement... air carrier accident liability insurance to protect the public from losses. This insurance information...

Estimating the cost of production stoppage

NASA Technical Reports Server (NTRS)

Delionback, L. M.

1979-01-01

Estimation model considers learning curve quantities, and time of break to forecast losses due to break in production schedule. Major parameters capable of predicting costs are number of units made prior to production sequence, length of production break, and slope of learning curve produced prior to break.

A simplified DEM-CFD approach for pebble bed reactor simulations

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

Li, Y.; Ji, W.

In pebble bed reactors (PBR's), the pebble flow and the coolant flow are coupled with each other through coolant-pebble interactions. Approaches with different fidelities have been proposed to simulate similar phenomena. Coupled Discrete Element Method-Computational Fluid Dynamics (DEM-CFD) approaches are widely studied and applied in these problems due to its good balance between efficiency and accuracy. In this work, based on the symmetry of the PBR geometry, a simplified 3D-DEM/2D-CFD approach is proposed to speed up the DEM-CFD simulation without significant loss of accuracy. Pebble flow is simulated by a full 3-D DEM, while the coolant flow field is calculatedmore » with a 2-D CFD simulation by averaging variables along the annular direction in the cylindrical geometry. Results show that this simplification can greatly enhance the efficiency for cylindrical core, which enables further inclusion of other physics such as thermal and neutronic effect in the multi-physics simulations for PBR's. (authors)« less

High voltage power supply with modular series resonant inverters

DOEpatents

Dreifuerst, Gary R.; Merritt, Bernard T.

1995-01-01

A relatively small and compact high voltage, high current power supply for a laser utilizes a plurality of modules containing series resonant half bridge inverters. A pair of reverse conducting thyristors are incorporated in each series resonant inverter module such that the series resonant inverter modules are sequentially activated in phases 360.degree./n apart, where n=number of modules for n>2. Selective activation of the modules allows precise output control reducing ripple and improving efficiency. Each series resonant half bridge inverter module includes a transformer which has a cooling manifold for actively circulating a coolant such as water, to cool the transformer core as well as selected circuit elements. Conductors connecting and forming various circuit components comprise hollow, electrically conductive tubes such as copper. Coolant circulates through the tubes to remove heat. The conductive tubes act as electrically conductive lines for connecting various components of the power supply. Where it is desired to make electrical isolation breaks, tubes comprised of insulating material such as nylon are used to provide insulation and continue the fluid circuit.

High voltage power supply with modular series resonant inverters

DOEpatents

Dreifuerst, G.R.; Merritt, B.T.

1995-07-18

A relatively small and compact high voltage, high current power supply for a laser utilizes a plurality of modules containing series resonant half bridge inverters. A pair of reverse conducting thyristors are incorporated in each series resonant inverter module such that the series resonant inverter modules are sequentially activated in phases 360{degree}/n apart, where n=number of modules for n>2. Selective activation of the modules allows precise output control reducing ripple and improving efficiency. Each series resonant half bridge inverter module includes a transformer which has a cooling manifold for actively circulating a coolant such as water, to cool the transformer core as well as selected circuit elements. Conductors connecting and forming various circuit components comprise hollow, electrically conductive tubes such as copper. Coolant circulates through the tubes to remove heat. The conductive tubes act as electrically conductive lines for connecting various components of the power supply. Where it is desired to make electrical isolation breaks, tubes comprised of insulating material such as nylon are used to provide insulation and continue the fluid circuit. 11 figs.

Breaks in the 45S rDNA Lead to Recombination-Mediated Loss of Repeats.

PubMed

Warmerdam, Daniël O; van den Berg, Jeroen; Medema, René H

2016-03-22

rDNA repeats constitute the most heavily transcribed region in the human genome. Tumors frequently display elevated levels of recombination in rDNA, indicating that the repeats are a liability to the genomic integrity of a cell. However, little is known about how cells deal with DNA double-stranded breaks in rDNA. Using selective endonucleases, we show that human cells are highly sensitive to breaks in 45S but not the 5S rDNA repeats. We find that homologous recombination inhibits repair of breaks in 45S rDNA, and this results in repeat loss. We identify the structural maintenance of chromosomes protein 5 (SMC5) as contributing to recombination-mediated repair of rDNA breaks. Together, our data demonstrate that SMC5-mediated recombination can lead to error-prone repair of 45S rDNA repeats, resulting in their loss and thereby reducing cellular viability. Copyright © 2016 The Authors. Published by Elsevier Inc. All rights reserved.

MELCOR computer code manuals

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Quartz antenna with hollow conductor

DOEpatents

Leung, Ka-Ngo; Benabou, Elie

2002-01-01

A radio frequency (RF) antenna for plasma ion sources is formed of a hollow metal conductor tube disposed within a glass tube. The hollow metal tubular conductor has an internal flow channel so that there will be no coolant leakage if the outer glass tube of the antenna breaks. A portion of the RF antenna is formed into a coil; the antenna is used for inductively coupling RF power to a plasma in an ion source chamber. The antenna is made by first inserting the metal tube inside the glass tube, and then forming the glass/metal composite tube into the desired coil shape.

Aircraft Accident Prevention: Loss-of-Control Analysis

NASA Technical Reports Server (NTRS)

Kwatny, Harry G.; Dongmo, Jean-Etienne T.; Chang, Bor-Chin; Bajpai, Guarav; Yasar, Murat; Belcastro, Christine M.

2009-01-01

The majority of fatal aircraft accidents are associated with loss-of-control . Yet the notion of loss-of-control is not well-defined in terms suitable for rigorous control systems analysis. Loss-of-control is generally associated with flight outside of the normal flight envelope, with nonlinear influences, and with an inability of the pilot to control the aircraft. The two primary sources of nonlinearity are the intrinsic nonlinear dynamics of the aircraft and the state and control constraints within which the aircraft must operate. In this paper we examine how these nonlinearities affect the ability to control the aircraft and how they may contribute to loss-of-control. Examples are provided using NASA s Generic Transport Model.

ZPPR-20 phase D : a cylindrical assembly of polyethylene moderated U metal reflected by beryllium oxide and polyethylene.

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

Lell, R.; Grimm, K.; McKnight, R.

The Zero Power Physics Reactor (ZPPR) fast critical facility was built at the Argonne National Laboratory-West (ANL-W) site in Idaho in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-20D Benchmark Assembly was part of a series of cores built in Assembly 20 (References 1 through 3) of the ZPPR facility to provide data for developing a nuclear power source for space applications (SP-100). The assemblies were beryllium oxide reflected and had core fuel compositions containing enriched uranium fuel, niobium and rhenium. ZPPR-20 Phase C (HEU-MET-FAST-075) was built as the reference flight configuration.more » Two other configurations, Phases D and E, simulated accident scenarios. Phase D modeled the water immersion scenario during a launch accident, and Phase E (SUB-HEU-MET-FAST-001) modeled the earth burial scenario during a launch accident. Two configurations were recorded for the simulated water immersion accident scenario (Phase D); the critical configuration, documented here, and the subcritical configuration (SUB-HEU-MET-MIXED-001). Experiments in Assembly 20 Phases 20A through 20F were performed in 1988. The reference water immersion configuration for the ZPPR-20D assembly was obtained as reactor loading 129 on October 7, 1988 with a fissile mass of 167.477 kg and a reactivity of -4.626 {+-} 0.044{cents} (k {approx} 0.9997). The SP-100 core was to be constructed of highly enriched uranium nitride, niobium, rhenium and depleted lithium. The core design called for two enrichment zones with niobium-1% zirconium alloy fuel cladding and core structure. Rhenium was to be used as a fuel pin liner to provide shut down in the event of water immersion and flooding. The core coolant was to be depleted lithium metal ({sup 7}Li). The core was to be surrounded radially with a niobium reactor vessel and bypass which would carry the lithium coolant to the forward inlet plenum. Immediately inside the reactor vessel was a rhenium baffle which would act as a neutron curtain in the event of water immersion. A fission gas plenum and coolant inlet plenum were located axially forward of the core. Some material substitutions had to be made in mocking up the SP-100 design. The ZPPR-20 critical assemblies were fueled by 93% enriched uranium metal because uranium nitride, which was the SP-100 fuel type, was not available. ZPPR Assembly 20D was designed to simulate a water immersion accident. The water was simulated by polyethylene (CH{sub 2}), which contains a similar amount of hydrogen and has a similar density. A very accurate transformation to a simplified model is needed to make any of the ZPPR assemblies a practical criticality-safety benchmark. There is simply too much geometric detail in an exact model of a ZPPR assembly, particularly as complicated an assembly as ZPPR-20D. 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 will be described in a later section. First, Assembly 20D was modeled in full detail--every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from this model were converted to an RZ model. ZPPR Assembly 20D has been determined to be an acceptable criticality-safety benchmark experiment.« less

System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment

NASA Astrophysics Data System (ADS)

Usov, E. V.; Butov, A. A.; Dugarov, G. A.; Kudasov, I. G.; Lezhnin, S. I.; Mosunova, N. A.; Pribaturin, N. A.

2017-07-01

The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.

Aircraft accident report : in-flight loss of propeller blade forced landing, and collision with terrain Atlantic Southeast Airlines, Inc., Flight 529 Embraer EMB-120RT, N256AS Carrollton, Georgia August 21, 1995

DOT National Transportation Integrated Search

1995-08-21

This report explains the accident involving Atlantic Southeast Airlines flight 529, an EMB-120RT airplane, which experienced the loss of a propeller blade and crashed during an emergency landing near Carrollton, Georgia, on August 21, 1995. Safety is...

Commuting by bike in Belgium, the costs of minor accidents.

PubMed

Aertsens, Joris; de Geus, Bas; Vandenbulcke, Grégory; Degraeuwe, Bart; Broekx, Steven; De Nocker, Leo; Liekens, Inge; Mayeres, Inge; Meeusen, Romain; Thomas, Isabelle; Torfs, Rudi; Willems, Hanny; Int Panis, Luc

2010-11-01

Minor bicycle accidents are defined as "bicycle accidents not involving death or heavily injured persons, implying that possible hospital visits last less than 24 hours". Statistics about these accidents and related injuries are very poor, because they are mostly not reported to police, hospitals or insurance companies. Yet, they form a major share of all bicycle accidents. Official registrations underestimate the number of minor accidents and do not provide cost data, nor the distance cycled. Therefore related policies are hampered by a lack of accurate data. This paper provides more insight into the importance of minor bicycle accidents and reports the frequency, risk and resulting costs of minor bicycle accidents. Direct costs, including the damage to bike and clothes as well as medical costs and indirect costs such as productivity loss and leisure time lost are calculated. We also estimate intangible costs of pain and psychological suffering and costs for other parties involved in the accident. Data were collected during the SHAPES project using several electronic surveys. The weekly prospective registration that lasted a year, covered 1187 persons that cycled 1,474,978 km. 219 minor bicycle accidents were reported. Resulting in a frequency of 148 minor bicycle accidents per million kilometres. We analyzed the economic costs related to 118 minor bicycle accidents in detail. The average total cost of these accidents is estimated at 841 euro (95% CI: 579-1205) per accident or 0.125 euro per kilometre cycled. Overall, productivity loss is the most important component accounting for 48% of the total cost. Intangible costs, which in past research were mostly neglected, are an important burden related to minor bicycle accidents (27% of the total cost). Even among minor accidents there are important differences in the total cost depending on the severity of the injury. 2010 Elsevier Ltd. All rights reserved.

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

Burger, J.R.

Loss control, both as a phrase and a concept, isn't used very widely in the U.S. coal industry although a U.S. manufacturer has cut accidents 71% and increased productivity 40% using the system. Safety is a part of the loss control concept, but it goes beyond traditional accident and illness prevention to become management control of anything that can result in loss or property damage. This includes what ILCI calls incidents, that is, ''any undesired or unwanted event that could (or does) degrade the efficiency of the business operation.'' These incidents could be accidents, quality or production problems, or evenmore » security breaches (such as thefts). So while safety is always a basic element-loss control also includes absenteeism control, security, fire prevention and industrial hygiene, since they're all interrelated disciplines for reducing loss. A baseline evaluation is followed by recommendations and guidance in self-sustaining corrective measures. This program would cost about $3,500 the first year. Possibly this approach is not used in the U.S. because miners feel that with all the legislation and regulation of the industry no further program is needed.« less

Alcoholism and drug abuse--some legal issues for employers.

PubMed

Howard, G

1990-05-01

Three specific areas of the law concern employers faced with problems of addiction at the workplace. At common law an employer may be guilty of negligence where a person has suffered personal injuries or economic loss as a result of an act of negligence committed in the course of employment by an employee. An example would be an employee with a serious addiction to alcohol or drugs who caused an accident in the company car whilst on company business. Employers may also be guilty of a criminal offence for breach of a statutory duty. One such duty is to have a 'safe system of work'. Other statutory rights guarantee employees a right not to be unfairly dismissed and this includes employees with addiction problems. Lastly, employers must be careful not to break the contract of employment if, for example, an employee with an addiction problem were to be suspended from duty or have his company car withdrawn, even if this was a temporary measure only.

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

NASA Astrophysics Data System (ADS)

Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

2018-01-01

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Preliminary risks associated with postulated tritium release from production reactor operation

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

O'Kula, K.R.; Horton, W.H.

1988-01-01

The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is assessing the off-site risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Other sources of tritium in the reactor are less likely to contribute to off-site risk in non-fuel melting accident scenarios. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as .5 kg) yields an estimate of /approximately/1 per reactor year. The full moderator loss frequency is conservatively chosen as 5 /times/ 10/sup /minus/3/ per reactor year. Conditional consequences, determined with amore » version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 4 /times/ 10/sup /minus/8/ per reactor year within 16 km of the release point. The full moderator loss accident contributes about 75% of the evaluated risks. 13 refs., 4 figs., 5 tabs.« less

Ocular injuries from laser accidents

NASA Astrophysics Data System (ADS)

Sliney, David H.

1996-04-01

Ocular injuries resulting from exposure to laser beams are relatively uncommon since there is normally a low probability of a relatively small-diameter laser beam entering the pupil of an eye. This has been the accident experience to date with lasers used in the research laboratory and in industry. A review of the accident data suggests that at least one type of laser is responsible for the majority of accidental injuries that result in a visual loss in the exposed eye. This is the q-switched neodymium:YAG laser. Although a continuous-wave laser causes a thermal coagulation of tissue, a q-switched laser having a pulse of only nanoseconds duration disrupts tissue. A visible or near-infrared laser can be focused on the retina, resulting in a vitreous hemorrhage. Examples of laser ocular injuries will be presented. Despite macular injuries and an initially serious visual loss, the vision of many patients recovers surprisingly well. Others may have severe vision loss. Corneal injuries resulting from exposure to reflected laser energy in the far-infrared account for surprisingly few reported laser accidents. The explanation for this accident statistic is not really clear. However, with the increasing use of lasers operating at many new wavelengths in the ultraviolet, visible and infrared, the ophthalmologist may see more accidental injuries from lasers.

A comprehensive review of rollover accidents involving vehicles equipped with Electronic Stability Control (ESC) systems.

PubMed

Padmanaban, Jeya; Shields, Leland E; Scheibe, Robert R; Eyges, Vitaly E

2008-10-01

This study investigated 478 police accident reports from 9 states to examine and characterize rollover crashes involving ESC-equipped vehicles. The focus was on the sequence of critical events leading to loss of control and rollover, and the interactions between the accident, driver, and environment. Results show that, while ESC is effective in reducing loss of control leading to certain rollover crashes, its effectiveness is diminished in others, particularly when the vehicle departs the roadway or when environmental factors such as slick road conditions or driver factors such as speeding, distraction, fatigue, impairment, or overcorrection are present.

A Comprehensive Review of Rollover Accidents Involving Vehicles Equipped with Electronic Stability Control (ESC) Systems

PubMed Central

Padmanaban, Jeya; Shields, Leland E.; Scheibe, Robert R.; Eyges, Vitaly E.

2008-01-01

This study investigated 478 police accident reports from 9 states to examine and characterize rollover crashes involving ESC-equipped vehicles. The focus was on the sequence of critical events leading to loss of control and rollover, and the interactions between the accident, driver, and environment. Results show that, while ESC is effective in reducing loss of control leading to certain rollover crashes, its effectiveness is diminished in others, particularly when the vehicle departs the roadway or when environmental factors such as slick road conditions or driver factors such as speeding, distraction, fatigue, impairment, or overcorrection are present. PMID:19026219

Polonium problem in lead-bismuth flow target

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

Pankratov, D.V.; Yefimov, E.I.; Bugreev, M.I.

1996-06-01

Alpha-active polonium nuclides Po198 - Po210 are formed in a lead-bismuth target as results of reactions Bi{sup 209}(n,{gamma})Bi{sup 210} {yields} Po{sup 210}, Bi{sup 209}(p,xn)Po{sup 210} {yields} Po{sup 210 {minus} x} (x = 1-12), Pb{sup 208}({alpha},xn) {yields} Po{sup 210 {minus} x + 2} (x = 2-14). The most important nuclides are Po-210 (T{sub {1/2}}=138.4 day), Po-209 (T{sub {1/2}}=102 years) and Po-208 (T{sub {1/2}}=2.9 years). Polonium activity of the circuit for SINQ - conditions is about 15,000 Ci after 1-year operation. Polonium radiation hazard is connected with its output from the coolant and formation of aerosol and surface alpha-activity after the circuitmore » break-down for repair works or in accidents. One of the important issues of polonium removal system creation is containing and storing polonium removed. Its storage in solidified alkaline is not expedient because of secondary neutron formation as a result of ({alpha},n) - reaction on oxygen and sodium nucleus. The estimations carried out demonstrated that by polonium concentration {approx} 100 Ci/l neutron current on the container surface can reach {approx} 10{sup 4}n/(cm{sup 2}s). Concentration and storage of polonium in solidified lead-bisumth seems the most convenient. The calculations demonstrated that in a 100 l container 50,000 Ci of polonium can be stored (as much as 3 times more than 1-year polonium product in SINQ-conditions) under temperature in the container less than melting point of lead bismuth (the wall temperature is about 100{degrees}C).« less

Nonverbal Communication and Aircrew Coordination in Army Aviation: Annotated Bibliography

DTIC Science & Technology

2006-06-01

limited in-cockpit visibility resulting from night and night vision goggle (NVG) use. The resulting reductions in nonverbal communication may have an...Readiness Center (USACRC) investigates Army aircraft accidents, and has included combat losses in their investigations. Their work continues to reveal a high...incidence of crew coordination errors that contribute to loss of life and equipment. A USACRC spokesman stated that, "Eighty percent of our accidents

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

75 FR 52039 - Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-24

... modifications which would replace the existing source range and intermediate range excore detector systems with... excore detector systems with equivalent neutron monitoring systems. The new instrumentation will perform... the design earthquake, the double design earthquake, the Hosgri earthquake, and the loss-of-coolant...

Fatal accidents following changes in daylight savings time: the American experience.

PubMed

Varughese; Allen

2001-01-01

Objective: This study examines specific hypotheses that both sleep loss and behavioral changes occurring with the time shifts for Daylight Savings Time (DST) significantly effect the number of fatal traffic accidents in the United States of America.Background: It has been reported that there is a significant increase in the number of automobile accidents in the spring shift to DST due to the loss of 1 h of sleep. But the extra hour gained at night with the shift from DST in the fall has been variably reported to be associated with increases and decreases in the number of automobile accidents which may reflect either behavioral anticipation with an extended late night prior to the change or the benefit of extra sleep after the change.Methods: Data from 21 years of United States' fatal automobile accidents were gathered. The mean number of accidents on the days at the time of the shifts (Saturday, Sunday and Monday) was compared to the average of the corresponding mean number of accidents on the matching day of the weeks preceding and following the shift. This was repeated for each DST shift. The number of accidents for a particular shift was also correlated with the year of the accidents.Results: There was a significant increase in accidents for the Monday immediately following the spring shift to DST (t=1.92, P=0.034). There was also a significant increase in number of accidents on the Sunday of the fall shift from DST (P<0.002). No significant changes were observed for the other days. A significant negative correlation with the year was found between the number of accidents on the Saturdays and Sundays but not Mondays.Conclusions: The sleep deprivation on the Monday following shift to DST in the spring results in a small increase in fatal accidents. The behavioral adaptation anticipating the longer day on Sunday of the shift from DST in the fall leads to an increased number of accidents suggesting an increase in late night (early Sunday morning) driving when traffic related fatalities are high possibly related to alcohol consumption and driving while sleepy. Public health educators should probably consider issuing warnings both about the effects of sleep loss in the spring shift and possible behaviors such as staying out later, particularly when consuming alcohol in the fall shift. Sleep clinicians should be aware that health consequences from forced changes in the circadian patterns resulting from DST come not only from physiological adjustments but also from behavioral responses to forced circadian changes.

Investigation of Dispersed and Dispersed Annular (rivulet or Thin Film) Flow Phase Separation in Tees.

NASA Astrophysics Data System (ADS)

McCreery, Glenn Ernest

An experimental and analytical investigation of dispersed and dispersed-annular (rivulet or thin film) flow phase separation in tees has been successfully completed. The research was directed at, but is not specific to, determining flow conditions, following a loss of coolant accident, in the large rectangular passageways leading to vacuum buildings in the containment envelope of some CANDU nuclear reactors. The primary objectives of the research were to: (1) obtain experimental data to help formulate and test mechanistic analytical models of phase separation, and (2) develop the analytical models in computer programs which predict phase separation from upstream flow and pressure conditions and downstream and side branch pressure boundary conditions. To meet these objectives an air-water experimental apparatus was constructed, and consists of large air blowers attached to a long rectangular duct leading to a tee in the horizontal plane. A variety of phenomena was investigated including, for comparison with computer predictions, air streamlines and eddy boundary geometry, drop size spectra, macroscopic mass balances, liquid rivulet pathlines, and trajectories of drops of known size and velocity. Four separate computer programs were developed to analyze phase separation. Three of the programs are used sequentially to calculate dispersed mist phase separation in a tee. The fourth is used to calculate rivulet or thin film pathlines. Macroscopic mass balances are calculated from a summation of mass balances for drops with representative sizes (and masses) spaced across the drop size spectrum. The programs are tested against experimental data, and accurately predict gas flow fields, drop trajectories, rivulet pathlines and macroscopic mass balances. In addition to development of the computer programs, analysis was performed to specify the scaling of dispersed mist and rivulet or thin film flow, to investigate pressure losses in tees, and the inter-relationship of loss coefficients, contraction coefficients, and eddy geometry. The important transient effects of liquid storage in eddies were also analyzed.

The DNA damage checkpoint pathway promotes extensive resection and nucleotide synthesis to facilitate homologous recombination repair and genome stability in fission yeast.

PubMed

Blaikley, Elizabeth J; Tinline-Purvis, Helen; Kasparek, Torben R; Marguerat, Samuel; Sarkar, Sovan; Hulme, Lydia; Hussey, Sharon; Wee, Boon-Yu; Deegan, Rachel S; Walker, Carol A; Pai, Chen-Chun; Bähler, Jürg; Nakagawa, Takuro; Humphrey, Timothy C

2014-05-01

DNA double-strand breaks (DSBs) can cause chromosomal rearrangements and extensive loss of heterozygosity (LOH), hallmarks of cancer cells. Yet, how such events are normally suppressed is unclear. Here we identify roles for the DNA damage checkpoint pathway in facilitating homologous recombination (HR) repair and suppressing extensive LOH and chromosomal rearrangements in response to a DSB. Accordingly, deletion of Rad3(ATR), Rad26ATRIP, Crb2(53BP1) or Cdc25 overexpression leads to reduced HR and increased break-induced chromosome loss and rearrangements. We find the DNA damage checkpoint pathway facilitates HR, in part, by promoting break-induced Cdt2-dependent nucleotide synthesis. We also identify additional roles for Rad17, the 9-1-1 complex and Chk1 activation in facilitating break-induced extensive resection and chromosome loss, thereby suppressing extensive LOH. Loss of Rad17 or the 9-1-1 complex results in a striking increase in break-induced isochromosome formation and very low levels of chromosome loss, suggesting the 9-1-1 complex acts as a nuclease processivity factor to facilitate extensive resection. Further, our data suggest redundant roles for Rad3ATR and Exo1 in facilitating extensive resection. We propose that the DNA damage checkpoint pathway coordinates resection and nucleotide synthesis, thereby promoting efficient HR repair and genome stability. © The Author(s) 2014. Published by Oxford University Press.

SiC Composite for Fuel Structure Applications

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

Yueh, Ken

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureablemore » weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2. Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.« less

Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

NASA Astrophysics Data System (ADS)

Li, Bo-Shiuan

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission gas release. An optimum design is sought considering both thermal and mechanical models of this ceramic cladding with UO2 and advanced high density fuels.

Fieldale-Collinsville Middle School: Banishing Anonymity

ERIC Educational Resources Information Center

Principal Leadership, 2010

2010-01-01

It is no accident that the staff at Fieldale-Collinsville Middle School adopted a central tenet of "Breaking Ranks in the Middle"--to banish anonymity by creating a personalized learning environment for all of its students. The school was created six years ago when the four middle schools in Henry County, VA, were consolidated into two…

Smart tungsten alloys as a material for the first wall of a future fusion power plant

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch.; Rasinski, M.; Kreter, A.; Unterberg, B.; Coenen, J. W.; Du, H.; Mayer, J.; Garcia-Rosales, C.; Calvo, A.; Ordas, N.

2017-06-01

Tungsten is currently deemed as a promising plasma-facing material (PFM) for the future power plant DEMO. In the case of an accident, air can get into contact with PFMs during the air ingress. The temperature of PFMs can rise up to 1200 °C due to nuclear decay heat in the case of damaged coolant supply. Heated neutron-activated tungsten forms a volatile radioactive oxide which can be mobilized into the atmosphere. New self-passivating ‘smart’ alloys can adjust their properties to the environment. During plasma operation the preferential sputtering of lighter alloying elements will leave an almost pure tungsten surface facing the plasma. During an accident the alloying elements in the bulk are forming oxides thus protecting tungsten from mobilization. Good plasma performance and the suppression of oxidation are required for smart alloys. Bulk tungsten (W)-chroimum (Cr)-titanium (Ti) alloys were exposed together with pure tungsten (W) samples to the steady-state deuterium plasma under identical conditions in the linear plasma device PSI 2. The temperature of the samples was ~576 °C-715 °C, the energy of impinging ions was 210 eV matching well the conditions expected at the first wall of DEMO. Weight loss measurements demonstrated similar mass decrease of smart alloys and pure tungsten samples. The oxidation of exposed samples has proven no effect of plasma exposure on the oxidation resistance. The W-Cr-Ti alloy demonstrated advantageous 3-fold lower mass gain due to oxidation than that of pure tungsten. New yttrium (Y)-containing thin film systems are demonstrating superior performance in comparison to that of W-Cr-Ti systems and of pure W. The oxidation rate constant of W-Cr-Y thin film is 105 times less than that of pure tungsten. However, the detected reactivity of the bulk smart alloy in humid atmosphere is calling for a further improvement.

[School accidents--an epidemiological assessment of injury types and treatment effort].

PubMed

Kraus, R; Heiss, C; Alt, V; Schnettler, R

2006-10-01

Children and adolescents spend up to 50% of their time at school. The purpose of this study was to assess injury patterns with their treatment of school accidents in a Trauma Service of a German University Hospital and to compare these data to the literature. All school accidents from 01.07.1999 to 30.06.2004 were statistically analysed in a retrospective manner by chart review. There were 1399 school accidents treated in our department. Average age of the injured children was 11.8 years with a boy:girl ratio of 3:2. Almost 40% of the injuries occurred during school sport. The most frequently injured region was the upper extremity including the hand (36.8%). Distortion and contusion were the most frequent diagnoses of all injuries. 16% of the cases had to be treated surgically and/or under general anaesthesia and also a total of 16% of the patients had to be admitted to the hospital. It can be concluded for school facilities that special attention has to be paid during school sports activity and breaks because they account for most accidents. Traffic education may reduce severe injuries. For diagnosis and treatment of school accidents specific knowledge of the growing longbones of the upper extremity and the hand is important.

Crew State Monitoring and Line-Oriented Flight Training for Attention Management

NASA Technical Reports Server (NTRS)

Stephens, Chad; Harrivel, Angela; Prinzel, Lawrence; Comstock, Ray; Abraham, Nijo; Pope, Alan; Wilkerson, James; Kiggins, Daniel

2017-01-01

Loss of control - inflight (LOC-I) has historically represented the largest category of commercial aviation fatal accidents. A review of worldwide transport airplane accidents (2001-2010) indicated that loss of airplane state awareness (ASA) was responsible for the majority of the LOC-I fatality rate. The Commercial Aviation Safety Team (CAST) ASA study identified 12 major themes that were indicated across the ASA accident and incident events. One of the themes was crew distraction or ineffective attention management, which was found to be involved in all 18 events including flight crew channelized attention, startle/surprise, diverted attention, and/or confirmation bias. Safety Enhancement (SE)-211, "Training for Attention Management" was formed to conduct research to develop and assess commercial airline training methods and realistic scenarios that can address these attention-related human performance limitations. This paper describes NASA SE-211 research for new design approaches and validation of line-oriented flight training (LOFT). Recent accident and incident data suggests that Spatial Disorientation (SD) and Loss-of-Energy State Awareness (LESA) for transport category aircraft are becoming an increasingly prevalent safety concern in all domestic and international operations (Commercial Aviation Safety Team, 2014a). SD is defined as an erroneous perception of aircraft attitude that can lead directly to a Loss-of-Control Inflight (LOC-I) event and result in an accident or incident. LESA is typically characterized by a failure to monitor or understand energy state indications (e.g., airspeed, altitude, vertical speed, commanded thrust) and a resultant failure to maintain safe flight.

H-division quarterly report, October--December 1977. [Lawrence Livermore Laboratory

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

Not Available

1978-02-10

The Theoretical EOS Group develops theoretical techniques for describing material properties under extreme conditions and constructs equation-of-state (EOS) tables for specific applications. Work this quarter concentrated on a Li equation of state, equation of state for equilibrium plasma, improved ion corrections to the Thomas--Fermi--Kirzhnitz theory, and theoretical estimates of high-pressure melting in metals. The Experimental Physics Group investigates properties of materials at extreme conditions of pressure and temperature, and develops new experimental techniques. Effort this quarter concerned the following: parabolic projectile distortion in the two-state light-gas gun, construction of a ballistic range for long-rod penetrators, thermodynamics and sound velocities inmore » liquid metals, isobaric expansion measurements in Pt, and calculation of the velocity--mass profile of a jet produced by a shaped charge. Code development was concentrated on the PELE code, a multimaterial, multiphase, explicit finite-difference Eulerian code for pool suppression dynamics of a hypothetical loss-of-coolant accident in a nuclear reactor. Activities of the Fluid Dynamics Group were directed toward development of a code to compute the equations of state and transport properties of liquid metals (e.g. Li) and partially ionized dense plasmas, jet stability in the Li reactor system, and the study and problem application of fluid dynamic turbulence theory. 19 figures, 5 tables. (RWR)« less

Bypass flow computations on the LOFA transient in a VHTR

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

Tung, Yu-Hsin; Johnson, Richard W.; Ferng, Yuh-Ming

2014-01-01

Bypass flow in the prismatic gas-cooled very high temperature reactor (VHTR) is not intentionally designed to occur, but is present in the gaps between graphite blocks. Previous studies of the bypass flow in the core indicated that the cooling provided by flow in the bypass gaps had a significant effect on temperature and flow distributions for normal operating conditions. However, the flow and heat transports in the core are changed significantly after a Loss of Flow Accident (LOFA). This study aims to study the effect and role of the bypass flow after a LOFA in terms of the temperature andmore » flow distributions and for the heat transport out of the core by natural convection of the coolant for a 1/12 symmetric section of the active core which is composed of images and mirror images of two sub-region models. The two sub-region models, 9 x 1/12 and 15 x 1/12 symmetric sectors of the active core, are employed as the CFD flow models using computational grid systems of 70.2 million and 117 million nodes, respectively. It is concluded that the effect of bypass flow is significant for the initial conditions and the beginning of LOFA, but the bypass flow has little effect after a long period of time in the transient computation of natural circulation.« less

An analytical investigation of transient effects on rewetting of heated thin flat plates

NASA Technical Reports Server (NTRS)

Platt, J. A.

1993-01-01

The rewetting of a hot surface is a problem of prime importance in the microgravity application of heat pipe technology, where rewetting controls the time before operations can be re-established following depriming of a heat pipe. Rewetting is also important in the nuclear industry (in predicting behavior during loss-of-coolant accidents), as well as in the chemical and petrochemical industries. Recently Chan and Zhang have presented a closed-form solution for the determination of the rewetting speed of a liquid film flowing over a finite (but long) hot plate subject to uniform heating. Unfortunately, their physically unreasonable initial conditions preclude a meaningful analysis of start-up transient behavior. A new nondimensionalization and closed-form solution for an infinitely-long, uniformly-heated plate is presented. Realistic initial conditions (step change in temperature across the wetting front) and boundary conditions (no spatial temperature gradients infinitely far from the wetting front) are employed. The effects of parametric variation on the resulting simpler closed-form solution are presented and compared with the predictions of a 'quasi-steady' model. The time to reach steady-state rewetting is found to be a strong function of the initial dry-region plate temperature. For heated plates it is found that in most cases the effect of the transient response terms cannot be neglected, even for large times.

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

Study of clad ballooning and rupture behaviour of Indian PHWR fuel pins under transient heating condition in steam environment

NASA Astrophysics Data System (ADS)

Sawarn, Tapan K.; Banerjee, Suparna; Sheelvantra, Smita S.; Singh, J. L.; Bhasin, Vivek

2017-11-01

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3-70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & ∼979 °C and a minimum at the burst temperature of ∼868 °C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon. A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data.

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

PubMed

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

Sensitivity to VSR failure: K pipe break accident

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

Meichle, R.H.

1969-09-12

Reactor effects of failure of a safety rod to scram can be considered in two major respects: The reduction in total safety system strength which will affect the amount of ``prompt drop`` and subsequent flux decay rate of the average neutron flux-level; and the change in local flux distribution due to the absence of the particular rod which fails to enter the reactor. The purpose of this memorandum is to describe the physical effects involved and to indicate the approximate magnitude of both reactor-wide and localized changes in event of failure of a VSR simultaneous with a K Reactor risermore » accident.« less

Parents' Responses to the Death of Adult Children from Accidents and Cancer: A Comparison.

ERIC Educational Resources Information Center

Shanfield, Stephen B.; And Others

1987-01-01

Compared parents (N=40) whose adult children died in traffic accidents to parents (N=24) whose adult children died of cancer. Cancer parents tended to experience loss less painfully than did accident parents. Differences between groups were explained by older age of children at death and less intense expression of grief. Circumstances of death…

Thrust performance of a variable-geometry, divergent exhaust nozzle on a turbojet engine at altitude

NASA Technical Reports Server (NTRS)

Straight, D. M.; Collom, R. R.

1983-01-01

A variable geometry, low aspect ratio, nonaxisymmetric, two dimensional, convergent-divergent exhaust nozzle was tested at simulated altitude on a turbojet engine to obtain baseline axial, dry thrust performance over wide ranges of operating nozzle pressure ratios, throat areas, and internal expansion area ratios. The thrust data showed good agreement with theory and scale model test results after the data were corrected for seal leakage and coolant losses. Wall static pressure profile data were also obtained and compared with one dimensional theory and scale model data. The pressure data indicate greater three dimensional flow effects in the full scale tests than with models. The leakage and coolant penalties were substantial, and the method to determine them is included.

Could driving safety be compromised by noise exposure at work and noise-induced hearing loss?

PubMed

Picard, Michel; Girard, Serge André; Courteau, Marilène; Leroux, Tony; Larocque, Richard; Turcotte, Fernand; Lavoie, Michel; Simard, Marc

2008-10-01

A study was conducted to verify if there is an association between occupational noise exposure, noise-induced hearing loss and driving safety expanding on previous findings by Picard, et al. (2008) that the two factors did increase accident risk in the workplace. This study was made possible when driving records of all Quebec drivers were made available by the Societe de l'assurance automobile du Quebec (SAAQ is the state monopoly responsible for the provision of motor vehicle insurance and the compensation of victims of traffic accidents). These records were linked with personal records maintained by the Quebec National Institute of Public Health as part of its mission to prevent noise induced hearing loss in the workplace. Individualized information on occupational noise exposure and hearing sensitivity was available for 46,030 male workers employed in noisy industries who also held a valid driver's permit. The observation period is of five years duration, starting with the most recent audiometric examination. The associations between occupational noise exposure levels, hearing status, and personal driving record were examined by log-binomial regression on data adjusted for age and duration of exposure. Daily noise exposures and bilateral average hearing threshold levels at 3, 4, and 6 kHz were used as independent variables while the dependent variables were 1) the number of motor vehicle accidents experienced by participants during the study period and 2) participants' records of registered traffic violations of the highway safety code. The findings are reported as prevalence ratios (PRs) with their 95% confidence intervals (CIs). Attributable numbers of events were computed with the relevant PRs, lesser-noise, exposed workers and those with normal hearing levels making the group of reference. Adjusting for age confirmed that experienced workers had fewer traffic accidents. The data show that occupational noise exposure and hearing loss have the same effect on driving safety record than that reported on the risk of accident in noisy industrial settings. Specifically, the risk of traffic accident (PR = 1.07 (CI 95% [1.01; 1.15]) is significantly associated with the daily occupational noise exposures >or= 100 dBA. For participants having a bilateral average hearing loss ranging from 16 to 30 dB, the PR of traffic accident is 1.06 (CI 95% [1.01; 1.11]) and reaches 1.31 (CI 95% [1.2; 1.42]) when the hearing loss exceeds of 50 dB. A reduction in the number of speeding violations occurred among workers occupationally exposed to noise levels >or= 90 dBA and those with noise-induced hearing loss >or=16 dB. By contrast, the same individuals had an increase in other violations of the Highway safety code. This suggests that noise-exposed workers might be less vigilant to other traffic hazards. Daily occupational noise exposures >or= 100 dBA and noise-induced hearing losses-even when just barely noticeable-may interfere with the safe operation of motor vehicles.

41 CFR 102-117.280 - What aspects of the TSP's performance are important to measure?

Code of Federal Regulations, 2010 CFR

2010-07-01

... (accidents, losses, damages or misdirected shipments) as a percentage of all shipments; (i) TSP's driving record (accidents, traffic tickets and driving complaints) as a percentage of shipments; and (j...

[The medical organizational aspects of decreasing of preventable mortality in the case of traffic accident in municipal district].

PubMed

Voloshina, L V; Plutnitskiĭ, A N

2010-01-01

The article deals with the results of the study of such actual issue as decreasing of preventable mortality in the case of traffic accident in municipal district. The analysis was based on the mortality statistical data and the expertise of causes of lethal outcomes of traffic accidents. The results are used to develop the measures of improving the organization and quality of medical care of victims of road accident on the pre-hospital and hospital stages on the level of municipal health care to decrease the human losses caused by traffic accident.

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.

Work time control, sleep & accident risk: A prospective cohort study.

PubMed

Tucker, Philip; Albrecht, Sophie; Kecklund, Göran; Beckers, Debby G J; Leineweber, Constanze

We examined whether the beneficial impact of work time control (WTC) on sleep leads to lower accident risk, using data from a nationally representative survey conducted in Sweden. Logistic regressions examined WTC in 2010 and 2012 as predictors of accidents occurring in the subsequent 2 years (N = 4840 and 4337, respectively). Sleep disturbance and frequency of short sleeps in 2012 were examined as potential mediators of the associations between WTC in 2010 and subsequent accidents as reported in 2014 (N = 3636). All analyses adjusted for age, sex, education, occupational category, weekly work hours, shift work status, job control and perceived accident risk at work. In both waves, overall WTC was inversely associated with accidents (p = 0.048 and p = 0.038, respectively). Analyses of the sub-dimensions of WTC indicated that Control over Daily Hours (influence over start and finish times, and over length of shift) did not predict accidents in either wave, while Control over Time-off (CoT; influence over taking breaks, running private errands during work and taking paid leave) predicted fewer accidents in both waves (p = 0.013 and p = 0.010). Sleep disturbance in 2012 mediated associations between WTC/CoT in 2010 and accidents in 2014, although effects' sizes were small (effectWTC = -0.006, 95% confidence interval [CI] = -0.018 to -0.001; effectCoT = -0.009, 95%CI = -0.022 to -0.001; unstandardized coefficients), with the indirect effects of sleep disturbance accounting for less than 5% of the total direct and indirect effects. Frequency of short sleeps was not a significant mediator. WTC reduces the risk of subsequently being involved in an accident, although sleep may not be a strong component of the mechanism underlying this association.

Simplified failure sequence evaluation of reactor pressure vessel head corroding in-core instrumentation assembly

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

McVicker, J.P.; Conner, J.T.; Hasrouni, P.N.

1995-11-01

In-Core Instrumentation (ICI) assemblies located on a Reactor Pressure Vessel Head have a history of boric acid leakage. The acid tends to corrode the nuts and studs which fasten the flanges of the assembly, thereby compromising the assembly`s structural integrity. This paper provides a simplified practical approach in determining the likelihood of an undetected progressing assembly stud deterioration, which would lead to a catastrophic loss of reactor coolant. The structural behavior of the In-Core Instrumentation flanged assembly is modeled using an elastic composite section assumption, with the studs transmitting tension and the pressure sealing gasket experiencing compression. Using the abovemore » technique, one can calculate the flange relative deflection and the consequential coolant loss flow rate, as well as the stress in any stud. A solved real life example develops the expected failure sequence and discusses the exigency of leak detection for safe shutdown. In the particular case of Calvert Cliffs Nuclear Power Plant (CCNPP) it is concluded that leak detection occurs before catastrophic failure of the ICI flange assembly.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

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

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

A damage assessment model of oil spill accident combining historical data and satellite remote sensing information: a case study in Penglai 19-3 oil spill accident of China.

PubMed

Wei, Lai; Hu, Zhuowei; Dong, Lin; Zhao, Wenji

2015-02-15

Oil spills are one of the major sources of marine pollution; it is important to conduct comprehensive assessment of losses that occur as a result of these events. Traditional methods are required to assess the three parts of losses including cleanup, socioeconomic losses, and environmental costs. It is relatively slow because assessment is complex and time consuming. A relatively quick method was developed to improve the efficiency of assessment, and then applied to the Penglai 19-3 accident. This paper uses an SAR image to calculate the oil spill area through Neural Network Classification, and uses historical oil-spill data to build the relationship between loss and other factors including sea-surface wind speed, and distance to the coast. A multiple regression equation was used to assess oil spill damage as a function of the independent variables. Results of this study can be used for regulating and quickly dealing with oil spill assessment. Copyright © 2014 The Authors. Published by Elsevier Ltd.. All rights reserved.

Mitigating the consequences of accidents

DOT National Transportation Integrated Search

2005-08-01

If an accident occurs, effective strategies must be available to minimize the effects in terms of loss of life, injury, and property damage. The Volpe Center's work in this area is reviewed in this report. It ranges from occupant protection strategie...

Biosocial profile of New Zealand prosthetic eye wearers.

PubMed

Pine, Keith R; Sloan, Brian; Jacobs, Robert J

2012-10-12

To describe the biosocial profile of New Zealand (NZ) artificial eye wearers and establish a basis for future research and international comparison. This retrospective study surveyed 431 NZ artificial eye wearers to investigate their ethnicity, gender, age, causes of eye loss, age of current prosthesis, ocular prosthetic maintenance regimes and the extent and severity of discharge associated with prosthesis wear. Approximately 3000 people wear artificial eyes in NZ. Accidents were the main cause of eye loss prior to 1990 and medical conditions have been the main cause since. In the 1960s, the ratio of men to women losing an eye from accidents was 5:1, but during the past decade the ratio was 1.4:1. Socket discharge occurred at least twice daily for one-third of the study group. Approximately 1 in 1440 people wear artificial eyes in NZ. Decline of eye loss due to accidents is consistent with decreasing workplace and traffic accidents and may be due to improved medical management, workplace safety standards and safer roads. Mucoid discharge is prevalent in the anophthalmic population of NZ and an evidence based treatment protocol for discharge associated with prosthesis wear is needed. Research into this distressing condition is planned.

Developing strategies for maintaining tank car integrity during train accidents

DOT National Transportation Integrated Search

2007-09-11

Accidents that lead to rupture of tank cars carrying : hazardous materials can cause serious public safety hazards and : substantial economic losses. The desirability of improved tank : car designs that are better equipped to keep the commodity : con...

U.S. Civil Rotorcraft Accidents, 1963 Through 1997

NASA Technical Reports Server (NTRS)

Harris, Franklin D.; Kasper, Eugene F.; Iseler, Laura E.

2000-01-01

Narrative summary data produced by the U.S. National Transportation Safety Board (NTSB) has been obtained and analyzed for all 8,436 U.S. civil registered rotorcraft accidents which occurred from mid-1963 through 1997. This analysis was based on the NTSB's assignment of each mishap into one of 21 "first event" categories. The number of U.S. civil registered rotorcraft as recorded by the Federal Aviation Administration (FAA) for the same period has also been obtained. Taken together, these data indicate the civil rotorcraft accident rate (on a per 1,000 registered rotorcraft basis) has decreased by almost a factor of 10 (i.e., from 130 accidents per 1,000 rotorcraft in 1964 to 13.4 per 1,000 in 1997). Analysis of the mishap data indicates over 70% of the rotorcraft accidents were associated with one of the following four NTSB "first event" categories: 2408 Loss of engine power (28.5%); 1,322 In-flight collisions with objects (15.7%); 1,114 Loss of control (13.2%); 1,083 Airframe/component/system failure or malfunction (12.8%).

Validation and Verification (V&V) of Safety-Critical Systems Operating Under Off-Nominal Conditions

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.

2012-01-01

Loss of control (LOC) remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft LOC accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or more often in combination. Hence, there is no single intervention strategy to prevent these accidents. Research is underway at the National Aeronautics and Space Administration (NASA) in the development of advanced onboard system technologies for preventing or recovering from loss of vehicle control and for assuring safe operation under off-nominal conditions associated with aircraft LOC accidents. The transition of these technologies into the commercial fleet will require their extensive validation and verification (V&V) and ultimate certification. The V&V of complex integrated systems poses highly significant technical challenges and is the subject of a parallel research effort at NASA. This chapter summarizes the V&V problem and presents a proposed process that could be applied to complex integrated safety-critical systems developed for preventing aircraft LOC accidents. A summary of recent research accomplishments in this effort is referenced.

[Agricultural occupational accidents in the county of Ringkoebing. Local registration].

PubMed

Carstensen, O; Rasmussen, K; Lauritsen, J M

1999-12-06

The aim of the study was to obtain knowledge of accidents and working conditions related to farm accidents for purposes of a preventive intervention. The survey was a prospective study following a cohort of 393 farms in the county of Ringkoebing in West-Jutland, Denmark. Every farm in the study registered work activity and farmwork related incidents that required a break in work. Every farm reporting an incident was interviewed by phone. The owner himself and the part-time farmer had the highest injury rate per working hour. There was no difference between animal and field related work looking at the incidence rate per workhour, but the incidence rate whilst repairing machinery and buildings was five times higher and highly significant. The results indicate the existence of groups and areas of risk, where a preventive effort is needed. The study is followed by a randomised intervention study among 200 Danish farms.

An novel identification method of the environmental risk sources for surface water pollution accidents in chemical industrial parks.

PubMed

Peng, Jianfeng; Song, Yonghui; Yuan, Peng; Xiao, Shuhu; Han, Lu

2013-07-01

The chemical industry is a major source of various pollution accidents. Improving the management level of risk sources for pollution accidents has become an urgent demand for most industrialized countries. In pollution accidents, the released chemicals harm the receptors to some extent depending on their sensitivity or susceptibility. Therefore, identifying the potential risk sources from such a large number of chemical enterprises has become pressingly urgent. Based on the simulation of the whole accident process, a novel and expandable identification method for risk sources causing water pollution accidents is presented. The newly developed approach, by analyzing and stimulating the whole process of a pollution accident between sources and receptors, can be applied to identify risk sources, especially on the nationwide scale. Three major types of losses, such as social, economic and ecological losses, were normalized, analyzed and used for overall consequence modeling. A specific case study area, located in a chemical industry park (CIP) along the Yangtze River in Jiangsu Province, China, was selected to test the potential of the identification method. The results showed that there were four risk sources for pollution accidents in this CIP. Aniline leakage in the HS Chemical Plant would lead to the most serious impact on the surrounding water environment. This potential accident would severely damage the ecosystem up to 3.8 km downstream of Yangtze River, and lead to pollution over a distance stretching to 73.7 km downstream. The proposed method is easily extended to the nationwide identification of potential risk sources.

Suicide in Peacekeepers: Risk Factors for Suicide versus Accidental Death

ERIC Educational Resources Information Center

Thoresen, Siri; Mehlum, Lars

2006-01-01

To investigate risk factors for suicide in veterans of peacekeeping, 43 suicides and 41 fatal accidents in Norwegian peacekeepers (1978 to 1995) were compared in a psychological autopsy study. Mental health problems were the most important risk factor for suicide. Both living alone and the break-up of a love relationship contributed uniquely to…

Transcription and replication: breaking the rules of the road causes genomic instability.

PubMed

Poveda, Ana Maria; Le Clech, Mikael; Pasero, Philippe

2010-01-01

Replication and transcription machineries progress at high speed on the same DNA template, which inevitably causes traffic accidents. Problems are not only caused by frontal collisions between polymerases, but also by cotranscriptional R-loops. These RNA-DNA hybrids induce genomic instability by blocking fork progression and could be implicated in the development of cancer.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

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

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

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

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references. (auth)

Experimental studies of thermal and chemical interactions between oxide and silicide nuclear fuels with water

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

farahani, A.A.; Corradini, M.L.

Given some transient power/cooling mismatch is a nuclear reactor and its inability to establish the necessary core cooling, energetic fuel-coolant interactions (FCI`s commonly called `vapor explosions`) could occur as a result of the core melting and coolant contact. Although a large number of studies have been done on energetic FCI`s, very few experiments have been performed with the actual fuel materials postulated to be produced in severe accidents. Because of the scarcity of well-characterized FCI data for uranium allows in noncommercial reactors (cermet and silicide fuels), we have conducted a series of experiments to provide a data base for themore » foregoing materials. An existing 1-D shock-tube facility was modified to handle depleted radioactive materials (U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al). Our objectives have been to determine the effects of the initial fuel composition and temperature and the driving pressure (triggering) on the explosion work output, dynamic pressures, transient temperatures, and the hydrogen production. Experimental results indicate limited energetics, mainly thermal interactions, for these fuel materials as compared to aluminum where more chemical reactions occur between the molten aluminum and water.« less

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

30 CFR 57.22241 - Advance face boreholes (I-C mines).

Code of Federal Regulations, 2010 CFR

2010-07-01

...) Boreholes shall be drilled in such a manner to insure that the advancing face will not accidently break into... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Advance face boreholes (I-C mines). 57.22241... Standards for Methane in Metal and Nonmetal Mines Ventilation § 57.22241 Advance face boreholes (I-C mines...

The Causal Relationship between DNA Damage Induction in Bovine Lymphocytes and the Fukushima Nuclear Power Plant Accident

PubMed Central

Nakamura, Asako J.; Suzuki, Masatoshi; Redon, Christophe E.; Kuwahara, Yoshikazu; Yamashiro, Hideaki; Abe, Yasuyuki; Takahashi, Shintaro; Fukuda, Tomokazu; Isogai, Emiko; Bonner, William M.; Fukumoto, Manabu

2017-01-01

The Fukushima Daiichi Nuclear Power Plant (FNPP) accident, the largest nuclear incident since the 1986 Chernobyl disaster, occurred when the plant was hit by a tsunami triggered by the Great East Japan Earthquake on March 11, 2011. The subsequent uncontrolled release of radioactive substances resulted in massive evacuations in a 20-km zone. To better understand the biological consequences of the FNPP accident, we have been measuring DNA damage levels in cattle in the evacuation zone. DNA damage was evaluated by assessing the levels of DNA double-strand breaks in peripheral blood lymphocytes by immunocyto-fluorescence-based quantification of γ-H2AX foci. A greater than two-fold increase in the fraction of damaged lymphocytes was observed in all animal cohorts within the evacuation zone, and the levels of DNA damage decreased slightly over the 700-day sample collection period. While the extent of damage appeared to be independent of the distance from the accident site and the estimated radiation dose from radiocesium, we observed age-dependent accumulation of DNA damage. Thus, this study, which was the first to evaluate the biological impact of the FNPP accident utilizing the γ-H2AX assays, indicated the causal relation between high levels of DNA damage in animals living in the evacuation zone and the FNPP accident. PMID:28240558

The Causal Relationship between DNA Damage Induction in Bovine Lymphocytes and the Fukushima Nuclear Power Plant Accident.

PubMed

Nakamura, Asako J; Suzuki, Masatoshi; Redon, Christophe E; Kuwahara, Yoshikazu; Yamashiro, Hideaki; Abe, Yasuyuki; Takahashi, Shintaro; Fukuda, Tomokazu; Isogai, Emiko; Bonner, William M; Fukumoto, Manabu

2017-05-01

The Fukushima Daiichi Nuclear Power Plant (FNPP) accident, the largest nuclear incident since the 1986 Chernobyl disaster, occurred when the plant was hit by a tsunami triggered by the Great East Japan Earthquake on March 11, 2011. The subsequent uncontrolled release of radioactive substances resulted in massive evacuations in a 20-km zone. To better understand the biological consequences of the FNPP accident, we have been measuring DNA damage levels in cattle in the evacuation zone. DNA damage was evaluated by assessing the levels of DNA double-strand breaks in peripheral blood lymphocytes by immunocytofluorescence-based quantification of γ-H2AX foci. A greater than two-fold increase in the fraction of damaged lymphocytes was observed in all animal cohorts within the evacuation zone, and the levels of DNA damage decreased slightly over the 700-day sample collection period. While the extent of damage appeared to be independent of the distance from the accident site and the estimated radiation dose from radiocesium, we observed age-dependent accumulation of DNA damage. Thus, this study, which was the first to evaluate the biological impact of the FNPP accident utilizing the γ-H2AX assays, indicated the causal relation between high levels of DNA damage in animals living in the evacuation zone and the FNPP accident.

Estimating probable flaw distributions in PWR steam generator tubes

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

Gorman, J.A.; Turner, A.P.L.

1997-02-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regardingmore » uncertainties and assumptions in the data and analyses.« less

PT-symmetry breaking with divergent potentials: Lattice and continuum cases

NASA Astrophysics Data System (ADS)

Joglekar, Yogesh N.; Scott, Derek D.; Saxena, Avadh

2014-09-01

We investigate the parity- and time-reversal (PT-) symmetry breaking in lattice models in the presence of long-ranged, non-Hermitian, PT-symmetric potentials that remain finite or become divergent in the continuum limit. By scaling analysis of the fragile PT threshold for an open finite lattice, we show that continuum loss-gain potentials Vα(x)∝i|x|αsgn(x) have a positive PT-breaking threshold for α >-2, and a zero threshold for α ≤-2. When α <0 localized states with complex (conjugate) energies in the continuum energy band occur at higher loss-gain strengths. We investigate the signatures of PT-symmetry breaking in coupled waveguides, and show that the emergence of localized states dramatically shortens the relevant time scale in the PT-symmetry broken region.

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

[Post-traumatic coma and pre-traumatic memory].

PubMed

Malacrida, R; Piazza, J; Abraham, G

1990-01-01

Instead of thinking that it is impossible to enter in the internal world of a comatose patient, we are now put before a new and encouraging prospective, that of the possibility, even though minimal, of influencing the vital residual organisation of the patient and to induce him perhaps to accept again external stimulations, which previously were too intense. As loss of conscience often causes loss of memory, our intention was to examine the problem of memory loss in comatose patients after accidents. The analysis of 50 questionnaires distributed to trauma-patients awakening from a comatose state and interviews give clear indications that: 1) the patients remember absolutely nothing during the time of the coma; 2) in the majority of cases (34) the patients remember in the moment preceding the accident a clear autodestructive tendency especially if they were the cause of the accident; and 3) almost all patients (41) agree to have benefited greatly from the trauma itself and from its memory.

Loss of labour productivity caused by disease and health problems: what is the magnitude of its effect on Spain's economy?

PubMed

Oliva-Moreno, Juan

2012-10-01

The aim of this study is to estimate the economic impact of the non-medical costs of diseases and accidents in Spain. Its main premise sustains the idea that in addition to the number of deaths, the loss of quality of life and the pain suffered by patients and their family members as a result of diseases and accidents, there are other indicators that provide us with a better understanding of their socioeconomic impact. Our analysis provides estimates of the loss of labour productivity in Spain as a result of health problems in 2005. Our main finding suggests an estimated loss amounting to over 37,969 millions euros, of which 9,136 millions euros are due to premature deaths, 18,577 millions to permanent disability and 10,255 millions to temporary disability. The loss in labour productivity due to accidents and health problems was estimated to a figure equivalent to nearly 4.2% of the Gross Domestic Product of Spain in 2005. This study underscores the strong economic impact of non-medical costs of diseases. In addition, it stresses the need for better information systems for collecting data that is relevant to the topic at hand.

Lessons Learned from the Fukushima Nuclear Accident due to Tohoku Region Pacific Coast Earthquake

NASA Astrophysics Data System (ADS)

Miki, M.; Wada, M.; Takeuchi, N.

2012-01-01

On March 11 2011, Great Eastern Japan Earthquake hit Japan and caused the devastating damage. Fukushima Nuclear Power Station (NPS) also suffered damages and provided the environmental effect with radioactive products. The situation has been settled to some extent about two months after the accidents, and currently, the cooling of reactor is continuing towards settling the situation. Japanese NPSs are designed based on safety requirements and have multiple-folds of hazard controls. However, according to publicly available information, due to the lager-than-anticipated Tsunami, all the power supply were lost, which resulted in loss of hazard controls. Also, although nuclear power plants are equipped with system/procedure in case of loss of all controls, recovery was not made as planned in Fukushima NPSs because assumptions for hazard controls became impractical or found insufficient. In consequence, a state of emergency was declared. Through this accident, many lessons learned have been obtained from the several perspectives. There are many commonality between nuclear safety and space safety. Both industries perform thorough hazard assessments because hazards in both industries can result in loss of life. Therefore, space industry must learn from this accident and reconsider more robust space safety. This paper will introduce lessons learned from Fukushima nuclear accident described in the "Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety" [1], and discuss the considerations to establish more robust safety in the space systems. Detailed information of Fukushima Dai-ichi NPS are referred to this report.

77 FR 7184 - Entergy Nuclear Indian Point 2, LLC; Entergy Nuclear Operations, Inc.; Indian Point Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-10

...-FR- H.1, ``Response To Loss Of Secondary Heat Sink.'' The NRC does not consider implementing 2-FR-H.1 an OMA, as actions to establish reactor coolant system decay heat removal can be performed from the... status trees if Auxiliary Feed necessary to Pump Building. establish alternate secondary heat sink...

[Costs of hand emergencies].

PubMed

Raimbeau, G

2003-10-01

In France at the present time, there is no comprehensive registry of hand injuries. Three types of occurrences; motor vehicle accidents, work accidents, and accidents incident to activities of daily living, are covered by different types of insurance. It is the individual insurance companies, payers of the indemnification, who maintain registries of these accidents. Statistics on work accidents are very detailed and consistent, but they are oriented toward risk management. The aggregate cost of traumatic injuries to the hand is not known. Only large financial institutions are equipped to determine appropriate preventive measures and to establish premium rates based on loss experience. In 2001, hand injuries accounted for 27% of work accidents causing loss of work of at least 1 day. About 29.8% of these work accidents caused permanent partial impairment. About 17.7% of total days lost and 18.2% of the total costs of permanent impairment were due to hand injuries. In the system of compensation for work accidents, there is a major difference in the cost according to the severity of the impairment. If the permanent impairment is equal to or less than 9%, a lump sum payment is made, but if the permanent impairment is over 9%, the worker receives regular payments for the rest of his life. In 2000, the average cost of a work injury with partial permanent impairment of over 9% was [symbol: see text] 85,405, while the average cost of a lump sum settlement was only [symbol: see text] 1479, a ratio of 57 to 1. The compensation costs represent 80% of the cost of work accidents, while the cost of treatment, including all providers and institutions, makes up only 20% of the cost. Compensation for sequelae of accidents in the course of daily life is new for the insurance companies, although these accidents are frequent and often cause significant repercussions in the professional lives of victims because of the loss of hand function. Provision of optimal treatment for these traumatic injuries from the very first moment is the best strategy for the third party payers. Better results of treatment not only reduce the costs of compensation for permanent partial impairment, but also greatly diminish the psychological impact on the injured worker. In the future, the management of the costs of these work injuries will become a priority. Pressure from consumers, which is growing, will favor this trend through the intermediary of private insurance companies.

Dynamics Modeling and Simulation of Large Transport Airplanes in Upset Conditions

NASA Technical Reports Server (NTRS)

Foster, John V.; Cunningham, Kevin; Fremaux, Charles M.; Shah, Gautam H.; Stewart, Eric C.; Rivers, Robert A.; Wilborn, James E.; Gato, William

2005-01-01

As part of NASA's Aviation Safety and Security Program, research has been in progress to develop aerodynamic modeling methods for simulations that accurately predict the flight dynamics characteristics of large transport airplanes in upset conditions. The motivation for this research stems from the recognition that simulation is a vital tool for addressing loss-of-control accidents, including applications to pilot training, accident reconstruction, and advanced control system analysis. The ultimate goal of this effort is to contribute to the reduction of the fatal accident rate due to loss-of-control. Research activities have involved accident analyses, wind tunnel testing, and piloted simulation. Results have shown that significant improvements in simulation fidelity for upset conditions, compared to current training simulations, can be achieved using state-of-the-art wind tunnel testing and aerodynamic modeling methods. This paper provides a summary of research completed to date and includes discussion on key technical results, lessons learned, and future research needs.

Loss of Control Prevention and Recovery: Onboard Guidance, Control, and Systems Technologies

NASA Technical Reports Server (NTRS)

Belcastro, Christine M.

2012-01-01

Loss of control (LOC) is one of the largest contributors to fatal aircraft accidents worldwide. LOC accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. These LOC hazards include vehicle impairment conditions, external disturbances; vehicle upset conditions, and inappropriate crew actions or responses. Hence, there is no single intervention strategy to prevent these accidents. NASA previously defined a comprehensive research and technology development approach for reducing LOC accidents and an associated integrated system concept. Onboard technologies for improved situation awareness, guidance, and control for LOC prevention and recovery are needed as part of this approach. Such systems should include: LOC hazards effects detection and mitigation; upset detection, prevention and recovery; and mitigation of combined hazards. NASA is conducting research in each of these areas. This paper provides an overview of this research, including the near-term LOC focus and associated analysis, as well as preliminary flight system architecture.

IMMEDIATE MENTAL CONSEQUENCES OF THE GREAT EAST JAPAN EARTHQUAKE AND FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT ON MOTHERS EXPERIENCING MISCARRIAGE, ABORTION, AND STILLBIRTH: THE FUKUSHIMA HEALTH MANAGEMENT SURVEY.

PubMed

Yoshida-Komiya, Hiromi; Goto, Aya; Yasumura, Seiji; Fujimori, Keiya; Abe, Masafumi

2015-01-01

The Fukushima Pregnancy and Birth Survey was launched to monitor pregnant mothers' health after the Great East Japan Earthquake and Fukushima Daiichi Nuclear Power Plant (NPP) accident. Several lines of investigations have indicated that a disaster impacts maternal mental health with childbirth. However, there is no research regarding mental health of mothers with fetal loss after a disaster. In this report, we focus on those women immediately after the Great East Japan Earthquake and Fukushima NPP accident and discuss their support needs. Data regarding 61 miscarriages, 5 abortions, and 22 stillbirths were analyzed among the women who were pregnant at the time of the accident in the present study. We used a two-item case-finding instrument for depression screening, and compared the childbirth group with the fetal loss groups. We also analyzed mothers' opinions written as free-form text. Among the three fetal loss groups, the proportion of positive depression screens was significantly higher in the miscarriage and stillbirth group than in the childbirth group. Mothers' opinions were grouped into six categories, with pregnancy-related items being most common, especially in the miscarriage and stillbirth groups. A higher proportion of Fukushima mothers with fetal loss, especially those with miscarriage and stillbirth, had depressive symptoms compared to those who experienced childbirth. Health care providers need to pay close attention to this vulnerable group and respond to their concerns regarding the effects on their fertility.

Results of Buoyancy-gravity Effects in ITER Cable-in- Conduit Conductor with Dual Channel

NASA Astrophysics Data System (ADS)

Bruzzone, P.; Stepanov, B.; Zanino, R.; Richard, L. Savoldi

2006-04-01

The coolant in the ITER cable-in-conduit conductors (CICC) flows at significant higher speed in the central channel than in the strand bundle region due to the large difference of hydraulic impedance. When energy is deposited in the bundle region, e.g. by ac loss or radiation, the heat removal in vertically oriented dual channel CICC with the coolant flowing downward is affected by the reduced density of helium (buoyancy) in the bundle region, which is arising from the temperature gradient due to poor heat exchange between the two channels. At large deposited power, flow stagnation and back-flow can cause in the strand bundle area a slow temperature runaway eventually leading to quench. A new test campaign of the thermal-hydraulic behavior was carried out in the SULTAN facility on an instrumented section of the ITER Poloidal Field Conductor Insert (PFIS). The buoyancy-gravity effect was investigated using ac loss heating, with ac loss in the cable calibrated in separate runs. The extent of upstream temperature increase was explored over a broad range of mass flow rate and deposited power. The experimental behavior is partly reproduced by numerical simulations. The results from the tests are extrapolated to the likely operating conditions of the ITER Toroidal Field conductor with the inboard leg cooled from top to bottom and heat deposited by nuclear radiation from the burning plasma.

The Safety of Physically Disabled Drivers

PubMed Central

Ysander, Lars

1966-01-01

Four hundred and ninety-four disabled drivers, the majority with loss of function in the legs usually as the result of poliomyelitis or amputations, have been studied with respect to the frequency of traffic accidents and serious traffic offences during a 10-year period. Traffic accidents which may have been caused by their disability occurred in only three (0·6%) of the total number of drivers investigated. In all three cases the driver had loss of function in the right leg. A comparison was made between the investigation series and a control series identical as regards sex, age, and licence-holding period but with a shorter exposure to traffic than the investigation series. The frequency of traffic accidents amounted to 7·1% in both series, and the frequency of serious traffic offences was 12·2% in the investigation series and 14·8% in the control series. Disabled drivers are not an increased hazard in traffic. The compensatory technical modifications to the vehicle which are generally adopted appeared to be adequate. However, there was a relatively increased frequency of accidents among drivers with loss of function in the right leg or right arm. An improvement of the technical modifications applied in these cases might result in a further reduction of the road-safety risks. PMID:4223486

Engine restart aid

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

Fedewa, Andrew

A system is disclosed comprising an engine having coolant passages defined therethrough, a first coolant pump, and a first radiator. The system additionally comprises a second coolant pump, a second radiator, and a liquid-to-air heat exchanger configured to condition the temperature of intake air to the engine. The system further includes a coolant valve means. For a first configuration of the coolant valve means the first coolant pump is configured to urge coolant through the coolant passages in the engine and through the first radiator, and the second coolant pump is configured to urge coolant through the liquid-to-air heat exchangermore » and through the second radiator. For a second configuration of the coolant valve means the second coolant pump is configured to urge coolant through the coolant passages in the engine and through the liquid-to-air heat exchanger. A method for controlling the system is also disclosed.« less

Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

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

Karthikeyan, R.; Tellier, R. L.; Hebert, A.

2006-07-01

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« less

Loss Estimations due to Earthquakes and Secondary Technological Hazards

NASA Astrophysics Data System (ADS)

Frolova, N.; Larionov, V.; Bonnin, J.

2009-04-01

Expected loss and damage assessment due to natural and technological disasters are of primary importance for emergency management just after the disaster, as well as for development and implementation of preventive measures plans. The paper addresses the procedures and simulation models for loss estimations due to strong earthquakes and secondary technological accidents. The mathematical models for shaking intensity distribution, damage to buildings and structures, debris volume, number of fatalities and injuries due to earthquakes and technological accidents at fire and chemical hazardous facilities are considered, which are used in geographical information systems assigned for these purposes. The criteria of technological accidents occurrence are developed on the basis of engineering analysis of past events' consequences. The paper is providing the results of scenario earthquakes consequences estimation and individual seismic risk assessment taking into account the secondary technological hazards at regional and urban levels. The individual risk is understood as the probability of death (or injuries) due to possible hazardous event within one year in a given territory. It is determined through mathematical expectation of social losses taking into account the number of inhabitants in the considered settlement and probability of natural and/or technological disaster.

How Past Loss of Control Accidents May Inform Safety Cases for Advanced Control Systems on Commercial Aircraft

NASA Technical Reports Server (NTRS)

Holloway, C. M.; Johnson, C. W.

2008-01-01

This paper describes five loss of control accidents involving commercial aircraft, and derives from those accidents three principles to consider when developing a potential safety case for an advanced flight control system for commercial aircraft. One, among the foundational evidence needed to support a safety case is the availability to the control system of accurate and timely information about the status and health of relevant systems and components. Two, an essential argument to be sustained in the safety case is that pilots are provided with adequate information about the control system to enable them to understand the capabilities that it provides. Three, another essential argument is that the advanced control system will not perform less safely than a good pilot.

Membrane-Based Gas Traps for Ammonia, Freon-21, and Water Systems to Simplify Ground Processing

NASA Technical Reports Server (NTRS)

Ritchie, Stephen M. C.

2003-01-01

Gas traps are critical for the smooth operation of coolant loops because gas bubbles can cause loss of centrifugal pump prime, interference with sensor readings, inhibition of heat transfer, and blockage of passages to remote systems. Coolant loops are ubiquitous in space flight hardware, and thus there is a great need for this technology. Conventional gas traps will not function in micro-gravity due to the absence of buoyancy forces. Therefore, clever designs that make use of adhesion and momentum are required for adequate separation, preferable in a single pass. The gas traps currently used in water coolant loops on the International Space Station are composed of membrane tube sets in a shell. Each tube set is composed of a hydrophilic membrane (used for water transport and capture of bubbles) and a hydrophobic membrane (used for venting of air bubbles). For the hydrophilic membrane, there are two critical pressures, the pressure drop and the bubble pressure. The pressure drop is the decrease in system pressure across the gas trap. The bubble pressure is the pressure required for air bubbles to pass across the water filled membrane. A significant difference between these pressures is needed to ensure complete capture of air bubbles in a single pass. Bubbles trapped by the device adsorb on the hydrophobic membrane in the interior of the hydrophilic membrane tube. After adsorption, the air is vented due to a pressure drop of approximately 1 atmosphere across the membrane. For water systems, the air is vented to the ambient (cabin). Because water vapor can also transport across the hydrophobic membrane, it is critical that a minimum surface area is used to avoid excessive water loss (would like to have a closed loop for the coolant). The currently used gas traps only provide a difference in pressure drop and bubble pressure of 3-4 psid. This makes the gas traps susceptible to failure at high bubble loading and if gas venting is impaired. One mechanism for the latter is when particles adhere to the hydrophobic membrane, promoting formation of a water layer about it that can blind the membrane for gas transport (Figure 1). This mechanism is the most probable cause for observed failures with the existing design. The objective of this project was to devise a strategy for choosing new membrane materials (database development and procedure), redesign of the gas trap to mitigate blinding effects, and to develop a design that can be used in ammonia and Freon-21 coolant loops.

Seveso 1986, Chernobyl 1976: a physicist' look at 2 ecological disasters

NASA Astrophysics Data System (ADS)

Ratti, S.

2004-05-01

Seveso suffered a chemical accident with a severe loss of supertoxic material (TCCD) released in the atmosphere; Chernobyl was a world known nuclear accident. The pollution induced by the two accident are analysed in term of fractal models. The first case involved a limited micro ecological system; the second one spread over a macro ecological system. The pollution is reproduced by means of simple Fractal Sum of Pulses models in the Seveso region; for the Chernobyl accident in northern Italy and in several european Countries. The 2 accidents are also analysed in terms of Universal Multifractals showing that thethe parameters α and C1 are those describing respectively rainfall (Seveso) and cloud formation (Chernobyl).

Cultural processes in psychotherapy for perinatal loss: Breaking the cultural taboo against perinatal grief.

PubMed

Markin, Rayna D; Zilcha-Mano, Sigal

2018-03-01

This paper argues that there is a cultural taboo against the public recognition and expression of perinatal grief that hinders parents' ability to mourn and their psychological adjustment following a loss. It is proposed that this cultural taboo is recreated within the therapy relationship, as feelings of grief over a perinatal loss are minimized or avoided by the therapist and parent or patient. Importantly, it is suggested that if these cultural dynamics are recognized within the therapy relationship, then psychotherapy has the immense opportunity to break the taboo by validating the parent's loss as real and helping the parent to mourn within an empathic and affect-regulating relationship. Specifically, it is suggested that therapists break the cultural taboo against perinatal grief and help parents to mourn through: acknowledging and not pathologizing perinatal grief reactions, considering intrapsychic and cultural factors that impact a parent's response to loss, exploring cultural reenactments within the therapy relationship, empathizing with the parent's experience of loss and of having to grieve within a society that does not recognize perinatal loss, coregulating the parent's feelings of grief and loss, and helping patients to create personally meaningful mourning rituals. Lastly, the impact of within and between cultural differences and therapist attitudes on the therapy process is discussed. (PsycINFO Database Record (c) 2018 APA, all rights reserved).

Directly connected heat exchanger tube section and coolant-cooled structure

DOEpatents

Chainer, Timothy J; Coico, Patrick A; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Steinke, Mark E

2014-04-01

A cooling apparatus for an electronics rack is provided which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures and a tube. The heat exchanger, which is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of distinct, coolant-carrying tube sections, each tube section having a coolant inlet and a coolant outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

CREW TRAINING - STS-33/51-L

NASA Image and Video Library

1985-09-12

S85-40171 (5 Sept. 1985) --- Astronaut Judith A. Resnik, in her office, at the Johnson Space Center (JSC). Resnik is taking a break from training for her upcoming space mission. EDITOR’S NOTE: The STS-51L crew members lost their lives in the space shuttle Challenger accident moments after launch on Jan. 28, 1986 from the Kennedy Space Center (KSC). Photo credit: NASA

Safety training priorities

NASA Astrophysics Data System (ADS)

Thompson, N. A.; Ruck, H. W.

1984-04-01

The Air Force is interested in identifying potentially hazardous tasks and prevention of accidents. This effort proposes four methods for determining safety training priorities for job tasks in three enlisted specialties. These methods can be used to design training aimed at avoiding loss of people, time, materials, and money associated with on-the-job accidents. Job tasks performed by airmen were measured using task and job factor ratings. Combining accident reports and job inventories, subject-matter experts identified tasks associated with accidents over a 3-year period. Applying correlational, multiple regression, and cost-benefit analysis, four methods were developed for ordering hazardous tasks to determine safety training priorities.

Introduction of Bayesian network in risk analysis of maritime accidents in Bangladesh

NASA Astrophysics Data System (ADS)

Rahman, Sohanur

2017-12-01

Due to the unique geographic location, complex navigation environment and intense vessel traffic, a considerable number of maritime accidents occurred in Bangladesh which caused serious loss of life, property and environmental contamination. Based on the historical data of maritime accidents from 1981 to 2015, which has been collected from Department of Shipping (DOS) and Bangladesh Inland Water Transport Authority (BIWTA), this paper conducted a risk analysis of maritime accidents by applying Bayesian network. In order to conduct this study, a Bayesian network model has been developed to find out the relation among parameters and the probability of them which affect accidents based on the accident investigation report of Bangladesh. Furthermore, number of accidents in different categories has also been investigated in this paper. Finally, some viable recommendations have been proposed in order to ensure greater safety of inland vessels in Bangladesh.

Cooling scheme for turbine hot parts

DOEpatents

Hultgren, Kent Goran; Owen, Brian Charles; Dowman, Steven Wayne; Nordlund, Raymond Scott; Smith, Ricky Lee

2000-01-01

A closed-loop cooling scheme for cooling stationary combustion turbine components, such as vanes, ring segments and transitions, is provided. The cooling scheme comprises: (1) an annular coolant inlet chamber, situated between the cylinder and blade ring of a turbine, for housing coolant before being distributed to the turbine components; (2) an annular coolant exhaust chamber, situated between the cylinder and the blade ring and proximate the annular coolant inlet chamber, for collecting coolant exhaust from the turbine components; (3) a coolant inlet conduit for supplying the coolant to said coolant inlet chamber; (4) a coolant exhaust conduit for directing coolant from said coolant exhaust chamber; and (5) a piping arrangement for distributing the coolant to and directing coolant exhaust from the turbine components. In preferred embodiments of the invention, the cooling scheme further comprises static seals for sealing the blade ring to the cylinder and flexible joints for attaching the blade ring to the turbine components.

Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core disruptive accident. Technical report 4

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

Romander, C. M.; Cagliostro, D. J.

Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-sec hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response.more » Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, an upper internals structure (UIS), and, in the more complex models SM 4 and SM 5, a Ni 200 thermal liner and core support structure. Water simulated the liquid sodium coolant and a low-density explosive simulated the HCDA loads.« less

The Analysis of the Contribution of Human Factors to the In-Flight Loss of Control Accidents

NASA Technical Reports Server (NTRS)

Ancel, Ersin; Shih, Ann T.

2012-01-01

In-flight loss of control (LOC) is currently the leading cause of fatal accidents based on various commercial aircraft accident statistics. As the Next Generation Air Transportation System (NextGen) emerges, new contributing factors leading to LOC are anticipated. The NASA Aviation Safety Program (AvSP), along with other aviation agencies and communities are actively developing safety products to mitigate the LOC risk. This paper discusses the approach used to construct a generic integrated LOC accident framework (LOCAF) model based on a detailed review of LOC accidents over the past two decades. The LOCAF model is comprised of causal factors from the domain of human factors, aircraft system component failures, and atmospheric environment. The multiple interdependent causal factors are expressed in an Object-Oriented Bayesian belief network. In addition to predicting the likelihood of LOC accident occurrence, the system-level integrated LOCAF model is able to evaluate the impact of new safety technology products developed in AvSP. This provides valuable information to decision makers in strategizing NASA's aviation safety technology portfolio. The focus of this paper is on the analysis of human causal factors in the model, including the contributions from flight crew and maintenance workers. The Human Factors Analysis and Classification System (HFACS) taxonomy was used to develop human related causal factors. The preliminary results from the baseline LOCAF model are also presented.

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

7 CFR 3570.67 - Project selection priorities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... emergencies, such as the loss of a community facility due to an accident or natural disaster or the loss of... funds, emergency conditions caused by economic problems, natural disasters, or leveraging of funds...

7 CFR 3570.67 - Project selection priorities.

Code of Federal Regulations, 2013 CFR

2013-01-01

... emergencies, such as the loss of a community facility due to an accident or natural disaster or the loss of... funds, emergency conditions caused by economic problems, natural disasters, or leveraging of funds...

7 CFR 3570.67 - Project selection priorities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... emergencies, such as the loss of a community facility due to an accident or natural disaster or the loss of... funds, emergency conditions caused by economic problems, natural disasters, or leveraging of funds...

Steam--water mixing and system hydrodynamics program. Task 4. Quarterly progress report, October 1, 1977--December 31, 1977. [PWR

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

Carbiener, W.A.; Cudnik, R.A.; Dykhuizen, R.C.

Experimental studies were conducted in a /sup 2///sub 15/-scale model of a four-loop pressurized water reactor at pressures to 75 psia to extend the understanding of steam-water interaction phenomena and processes associated with a loss-of-coolant accident. Plenum filling studies were conducted with hydraulic communication between the cold leg and core steam supplies and hot walls, with both fixed and ramped steam flows. Comparisons of correlational fits have been made for penetration data obtained with hydraulic communication, fixed cold leg steam, and no cold leg steam. Statistical tests applied to these correlational fits have indicated that the hydraulic communication and fixedmore » cold leg steam data can be considered to be a common data set. Comparing either of these data sets to the no cold leg steam data using the statistical test indicated that it was unlikely that these sets could be considered to be a common data set. The introduction of cold leg steam results in a slight decrease in penetration relative to that obtained without cold leg steam at the same value of subcooling of water entering the downcomer. A dimensionless parameter which is a weighted mean of a modified Froude number and the Weber number has been proposed as a scaling parameter for penetration data. This parameter contains an additional degree of freedom which allows data from different scales to collapse more closely to a single curve than current scaling parameters permit.« less

Assessment of safety margins in zircaloy oxidation and embrittlement criteria for ECCS acceptance

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

Williford, R.E.

1986-04-01

Current Emergency Core Cooling System (ECCS) Acceptance Criteria for light-water reactors include certain requirements pertaining to calculations of core performance during a Loss of Coolant Accident (LOCA). The Baker-Just correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (17% ECR). The minimum margin of safety was estimated for each of these criteria, based on research performed in the last decade. Margins were defined as the amounts of conservatism over and above the expected extreme values computed from the data base atmore » specified confidence levels. The currently required Baker-Just oxidation correlation provides margins only over the 1100/sup 0/C to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperatures above 1204/sup 0/C for 210 and 160 seconds, respectively, at the 95% confidence level. ECR thermal shock and handling margins at the 50% and 95% confidence levels, respectively, range between 2% and 7% ECR for the Baker-Just correlation, but vanish at temperatures between 1100/sup 0/C and 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. Use of the Cathcart-Pawel correlation for LOCA calculations can be justified at the 85% to 88% confidence level if cooling rate effects can be neglected. 75 refs., 21 figs.« less

Directly connected heat exchanger tube section and coolant-cooled structure

DOEpatents

Chainer, Timothy J.; Coico, Patrick A.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

2015-09-15

A method is provided for fabricating a cooling apparatus for cooling an electronics rack, which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures, and a tube. The heat exchanger is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of coolant-carrying tube sections, each tube section having a coolant inlet and outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

Examination of Icing Induced Loss of Control and Its Mitigations

NASA Technical Reports Server (NTRS)

Reehorst, Andrew L.; Addy, Harold E., Jr.; Colantonio, Renato O.

2010-01-01

Factors external to the aircraft are often a significant causal factor in loss of control (LOC) accidents. In today s aviation world, very few accidents stem from a single cause and typically have a number of causal factors that culminate in a LOC accident. Very often the "trigger" that initiates an accident sequence is an external environment factor. In a recent NASA statistical analysis of LOC accidents, aircraft icing was shown to be the most common external environmental LOC causal factor for scheduled operations. When investigating LOC accident or incidents aircraft icing causal factors can be categorized into groups of 1) in-flight encounter with super-cooled liquid water clouds, 2) take-off with ice contamination, or 3) in-flight encounter with high concentrations of ice crystals. As with other flight hazards, icing induced LOC accidents can be prevented through avoidance, detection, and recovery mitigations. For icing hazards, avoidance can take the form of avoiding flight into icing conditions or avoiding the hazard of icing by making the aircraft tolerant to icing conditions. Icing detection mitigations can take the form of detecting icing conditions or detecting early performance degradation caused by icing. Recovery from icing induced LOC requires flight crew or automated systems capable of accounting for reduced aircraft performance and degraded control authority during the recovery maneuvers. In this report we review the icing induced LOC accident mitigations defined in a recent LOC study and for each mitigation describe a research topic required to enable or strengthen the mitigation. Many of these research topics are already included in ongoing or planned NASA icing research activities or are being addressed by members of the icing research community. These research activities are described and the status of the ongoing or planned research to address the technology needs is discussed

[Evaluation of regulatory policies: the prevention of traffic accidents in Spain].

PubMed

Villalbí, Joan R; Pérez, Catherine

2006-03-01

Traffic accident injuries may be reduced with public policies. We review regulatory policies extending beyond the health sector by studying the case of traffic accident injuries. They have been the object of other analyses in Spain by both health professionals and professionals from other sectors, but we have not found a previous thorough review including regulatory aspects. We analyze the evolution of fatal victims of traffic accidents as collected by the Dirección General de Tráfico, stratifying for pedestrians, two-wheel vehicle occupants and occupants of other vehicles, and breaking down accidents between those occurring in roads and in urban settings. Despite the increase in exposure factors between 1970 and 2003, we observe a strong impact of regulatory policies in accident mortality. A favorable impact is seen for regulations and enforcement actions on motorcycle helmets, speed limits and the control of alcohol use, and a lower impact for safety belts, perhaps because its actual effective implementation was not equally sharp. The adoption of comprehensive plans or complex legislation packages seems to have had a positive impact, perhaps attributable to its triggering of more effective enforcement of already existing regulations. Although the existence of legal norms is not enough in itself, as its impact is low without active enforcement, compliance improves over time. In any case, the existence of specific initiatives to influence this field is important to obtain the best results of regulatory policies in public health.

Containment Sodium Chemistry Models in MELCOR.

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

Louie, David; Humphries, Larry L.; Denman, Matthew R

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRCmore » code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.« less

How safe are HEMS-programmes in Germany? A retrospective analysis.

PubMed

Thies, Karl-Christian; Sep, Daan; Derksen, Remon

2006-03-01

Recent accidents with helicopter emergency medical service (HEMS) aircraft raise the question how safe HEMS in Germany is and how accidents could be prevented. We surveyed all German HEMS-programmes and reviewed the data of the German Aviation Authority regarding accidents with HEMS. An average German HEMS-programme encounters one accident leading to at least severe damage or loss of the helicopter in 26 operating years, one accident resulting in casualties in 65 operating years and one fatal accident in 111 operating years. The major causes of accidents were obstacle strikes during landing at the scene. Flying in bad weather conditions and lack of discipline were other factors contributing to HEMS-accidents. HEMS-safety could be improved by special training programmes for pilots and HEMS-crewmembers to address the factors listed above. Safety training for doctors is recommended but we did not find support for the notion of changing the doctor's legal position of a passenger to a HEMS-crewmember.

The physiotherapeutic context of loss of dominant arm function due to occupational accidents.

PubMed

Kostiukow, Anna; Kaluga, Elżbieta; Samborski, Włodzimierz; Rostkowska, Elżbieta

2016-12-23

The study examines the problem of dominant arm function loss in rural adult patients due to work-related accidents. The types of risks involved in farmyard work include falling from a height, manually moving loads, overturning/accident whilst driving an agricultural tractor, noise and vibration, use of pesticides, and the risk of being cut or injured. The study focuses on adaptation of the non-dominant arm. The main aim of the study was evaluation of visual-motor coordination on the basis of performance of the non-dominant hand in patients after the loss of function of the dominant arm. The research sample consisted of 52 patients with a permanent or temporary loss of function or severely limited function of the dominant arm. The subjects were patients with arm amputations due to various occupational injuries sustained while operating agricultural and construction machinery and forestry equipment, following traumas or complicated medical surgeries of the arm, or due to car accidents. The following tests were applied in the analysis: I) Dufour cross-shaped apparatus test for assessing visual motor-coordination; II) paper-and-pencil tests and the Relay Baton motor fitness test; III) anthropometric measurements; IV) Edinburgh Handedness Inventory; and V) a questionnaire survey. The results of the apparatus and motor tests indicate the same tendency: reaction to stimuli measured on the basis of performance of the non-dominant arm is longer in shorter and older patients. Visual-motor coordination, as measured by the performance of the non-dominant arm, is significantly affected by the subject's body height and arm length.